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The views expressed in this report are those of the authors and do not necessarily reflectthose of the European Commission

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1.1. OBJECTIVES1.2. GENERAL1.3. ORGANISATIONS INVOLVED IN THE PROJECT

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2.1. SYNTHESIS OF WORK DONE BY OECD-NEA, IAEA AND WANO2.1.1. OECD Nuclear Energy Agency2.1.2. IAEA2.1.3. WANO2.1.4. International Databases

2.2. SYNTHESIS OF WORK DONE UNDER EUROPEAN CONTEXT2.2.1. Introduction2.2.2. General orientation or European research and training programmes in the

field of nuclear energy2.2.3. The Working Group on Codes and Standards (WGCS)2.2.4. The European networks: NESC, AMES, ENIQ2.2.5. Conclusions

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3.1. INTRODUCTION

3.2. TECHNOLOGICAL ASPECTS3.2.1. Ageing of systems, structures and components3.2.2. Methods of identifying ageing in service3.2.3. Mitigation of ageing effects (prevent, delay, restore)

3.3. OTHER ASPECTS3.3.1. General3.3.2. Technological obsolescence3.3.3. Ageing of culture and procedure

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4.1. BELGIAN APPROACH4.1.1. Introduction4.1.2. General regulatory framework4.1.3. License life4.1.4. Design life

4.2. FRENCH APPROACH4.2.1. General regulatory framework4.2.2. The context in which ageing is monitored

4.3. SPANISH APPROACH4.3.1. General4.3.2. Regulatory requirements4.3.3. NPP lifetime management programmes4.3.4. Periodic safety reviews4.3.5. UNESA guideline for periodic safety review development

4.4. OTHER EUROPEAN COUNTRIES

4

4.4.1. General4.4.2. United Kingdom

4.5. COMPARISON OF REGULATORY APPROACHES4.5.1. General4.5.2. Periodic safety reviews4.5.3. Ageing management programs4.5.4. Generic evaluations

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5.1. BELGIAN APPROACH5.1.1. Introduction5.1.2. Systems, structures and components – Prioritisation of components5.1.3. Selection / identification of ageing mechanisms5.1.4. Ageing prediction criteria5.1.5. Surveillance / periodic testing5.1.6. Mitigation of ageing effects5.1.7. Maintenance programs / component repair / replacement / improvement

5.2. FRENCH APPROACH5.2.1. Introduction5.2.2. Systems, structures and components – Prioritisation of components5.2.3. Selection / identification of ageing mechanisms5.2.4. Ageing prediction criteria5.2.5. Surveillance / periodic testing5.2.6. Mitigation of ageing effects5.2.7. Maintenance programs / component repair / replacement / improvement5.2.8. Other considerations

5.3. SPANISH APPROACH5.3.1. Introduction5.3.2. Systems, structures and components – Prioritisation of components5.3.3. Selection / identification of ageing mechanisms5.3.4. Ageing prediction criteria5.3.5. Surveillance / periodic testing5.3.6. Mitigation of ageing effects5.3.7. Maintenance programs / component repair / replacement / improvement

5.4. COMPARISON OF UTILITIES APPROACHES5.4.1. Introduction5.4.2. Organisation5.4.3. Prioritisation of systems, structures and components5.4.4. Identification and selection of ageing mechanisms5.4.5. Management of ageing mechanisms5.4.6. Conclusions

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6.1. FOREWORD

6.2. DESCRIPTION OF APPROACHES BY COUNTRY

6.3. COMPARISON IN TERM OF STRATEGY6.3.1. Documentation6.3.2. Safety and regulatory aspects6.3.3. Ageing management programme policy / Scope of AMP

5

6.4. COMPARISON IN TERM OF ORGANISATION6.4.1. Differences due to the context of the companies6.4.2. Internal organisation of ageing management6.4.3. Management of personnel6.4.4. Expert appraisal / advice

6.5. COMPARISON IN TERM OF ACTIVITIES6.5.1. Methodologies6.5.2. Operating procedures6.5.3. Surveillance programme6.5.4. Maintaining equipment qualification

6.6. COMPARISON IN TERM OF MONITORING

6.7. CONCLUSION

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7.1. CONCLUSIONS7.2. RECOMMENDATIONS

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A. OECD NUCLEAR ENERGY AGENCY DOCUMENTSB. IAEA DOCUMENTS

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1. GENERAL1.1. Introduction1.2. Screening methodology1.3. Criteria definition1.4. Implementation

2. BELGIUM APPROACH

3. FRENCH APPROACH

4. SPANISH APPROACH

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1. BELGIUM EXAMPLES1.0. General1.1. Reactor pressure vessel1.2. Vessel internals1.3. Steam generator1.4. Primary pump1.5. Primary piping1.6. Fatigue1.7. Piping1.8. Reactor containment1.9. Electrical and I&C equipment1.10. Electrical cables

6

2. FRENCH EXAMPLES2.1. Reactor pressure vessel – Irradiation embrittlement2.2. Pipes, cast elbows and primary coolant pump casings2.3. Instrumentation and control2.4. Reactor containment2.5. Electrical cables

3. SPANISH EXAMPLES3.0. General3.1. Reactor pressure vessel – Irradiation embrittlement3.2. Steam generator3.3. Class 1 piping – Fatigue (thermal stratification)3.4. Piping – Erosion/corrosion3.5. Concrete structures3.6. Cables

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1. INTRODUCTION2. PURPOSE

3. COMPARATIVE STUDY OF THE METHODOLOGIES3.1. Guideline 95-10 (Rev. 1, January 2000)3.2. IAEA's methodology for ageing control in components important for

nuclear safety (Technical Report Series N°338)3.3. Life management methodology established by UNESA3.4. Comparison of the methodologies

4. CONCLUSIONS5. REFERENCES

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1. INTRODUCTION2. DOCUMENTS

7

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AG: Activity Groups of WGCSAMAT: Ageing Management Assessment Teams of IAEAAMES: Ageing Materials Evaluation and StudiesAMP: Ageing Management ProgrammeANSI: American National Standards InstituteASME: The American Society of Mechanical EngineersASTM: The American Society for Testing and MaterialsATWS: Anticipated Transient Without ScramAVB: Anti-Vibration Bars in Steam GeneratorsAVN: Association Vinçotte NuclearAVT: All Volatile TreatmentBWR: Boiling Water ReactorCDD: Component Degradation DatasheetCNRA: Committee on Nuclear Regulatory Activities of OECDCRC: Corrosion Resistant CladdingCRP: Co-ordinated Research Projects of IAEACRPPH: Committee on Radiation Protection and Public Health of NEACSN: Spanish Nuclear Safety CouncilCSNI: Committee on the Safety of Nuclear Installations of OECDDG: Directorates-General of the ECDNPA: Small Angles Neutron DiffractionDSIN: French Nuclear Installations Safety DirectorateEC: European CommissionECCS: Emergency Core Cooling SystemEDG: Emergency Diesels GeneratorsENIQ: European Network for Inspection QualificationENIS-G: European Nuclear Installations Safety GroupEPRI: Electric Power Research InstituteEU: European UnionFAC: Flow Accelerated CorrosionFBR: Fast Breeder ReactorFIM: French Average Irradiation embrittlement prediction formulaFIS: French Upper Bound Irradiation embrittlement prediction formulaFP: Euratom Framework ProgrammeFSAR: Final Safety Analysis ReportGALL:Generic Ageing Lessons LearnedHSE: British Health and Safety ExecutiveIAEA: International Atomic Energy AgencyIASCC: Irradiation Assisted Stress Corrosion CrackingICDE: International Common cause failure Data ExchangeIDPRVM: International Database on reactor Pressure Vessel MaterialsIEEE: The Institute of Electrical and Electronics EngineersIGSCC: Intergranular Stress Corrosion CrackingINSAG: International Nuclear Safety Advisory Group of IAEAISI: In-Service InspectionIWG-LMNPP: International Working Group on Life Management of Nuclear PowerPlantsJRC: Joint Research Center of the ECKTA: Kern Technische AusschussLCM: Lifetime Cycle ManagementLWR: Light Water Reactor

8

MIC: Microbiologically Influenced CorrosionMOV: Motor Operated ValvesMPDS: Maintenance Practices Data SheetMR: Maintenance RulesNDC: Nuclear Development Committee of OECDNDE: Non Destructive ExaminationNDT: Non Destructive Testing or Nil Ductility Temperature, depending on thecontextNEA: Nuclear Energy Agency of OECDNEI: US Nuclear Energy InstituteNII: Nuclear Installation Inspectorate of United KingdomNPP: Nuclear Power PlantNESC: Network for Evaluating Steel ComponentsNRC: United States Nuclear Regulatory CommissionNRWG: Nuclear Regulator’s Working Group of the ECNSSS: Nuclear Steam Supply SystemNUSS: Nuclear Safety Standards of IAEAOECD: Organisation for Economic Co-operation and DevelopmentPISC: Programme for the Inspection of Steel ComponentsPLEX: Plant Life ExtensionPLIM: Plant Life ManagementPSA: Probabilistic Safety AssessmentPSR: Periodic Safety ReviewPTE: Thermo-Electric PowerPWG: Principal Working Groups of CSNIPWR: Pressurised Water ReactorPWSCC: Primary Water Stress Corrosion CrackingRCC: French Design and Construction Rules for Components of PWR NuclearIslandsRG: Regulatory Guide of the US-NRCRPV: Reactor Pressure VesselRSE-M: French In Service Inspection Rules for Mech. Components of PWR

Nuclear IslandsRSK: German Reactors Safety CommissionRWMC: Radioactive Waste Management Committee of NEASCC: Stress Corrosion CrackingSESAR/FAP: Senior Group of Experts on Nuclear Safety Research Facilities andProgrammesSG: Safety Guides of IAEA, or Steam Generator, depending on the contextSSC: Systems, Structures and ComponentsTECDOC: Technical Document of IAEATGSCC: Transgranular Stress Corrosion CrackingTLAA:Time Limited Ageing AnalysesUS DOE: United States Department of EnergyVISA: French qualification procedure for electrical equipment (Ageing,

Irradiation, Seismic and Accident environment conditions)WANO: World Association of Nuclear OperatorsWGCS: Working Group on Codes and Standards of EC

9

�� ,1752'8&7,21

�� 2%-(&7,9(6The objectives of the study covered by this report are defined in [57]. They consist inproviding recommendations for the development of a methodology to monitor, controland anticipate the ageing of Nuclear Islands, in order to maintain their level of safetyduring the whole NPP life cycle.

This is obtained through a synthesis of the potential ageing mechanisms, their potentialeffects and the available identification and mitigation methods, and an evaluation of theexisting ageing management practices in Belgium, France and Spain, using internationalrecommendations as guidance documents.

The aspects covered include, in particular, classification of component priorities,identification of degradation phenomena, surveillance methods and preventivemaintenance and repair / replacement programs.

This synthesis comparison work is the basis for the recommendations issued at the endof the study.

�� *(1(5$/Ageing in nuclear power plants shall be managed so as to ensure that the designfunctions remain available throughout the service life of the plant. From the safetyperspective, this implies that ageing degradation of systems, structures and componentsimportant to safety remain within acceptable limits, and that procedures and personneltraining remain adapted.

Ageing means evolution of personnel and procedure adequacy and evolution of materialor equipment properties, which, after a certain time, may not be compatible with therequired safety margins, or with an economic functioning of the plant. Repair orreplacement of components, as well as change in service conditions for a bettercompatibility with component reduced capacities are possible.

In parallel, safety requirements may increase with time, following the evolution of thepublic acceptance, and costs of other energy sources may decrease, limiting as a resultthe economic interest of a continued operation. The plant life will then be the result ofthe consideration of ageing in a changing regulatory, political, technical and economiccontext.

In the context of stable safety requirements, the evolution of the safety margins for agiven component can be illustrated as shown in figure 1-1 [89]. This figure – which isgiven for illustrative purpose only and do not take into account the possibility ofcomponent replacement – shows that a better evaluation of the applied loading, as wellas a better knowledge of component performance can result in a modification of thedemonstrated safety margin, and consequently of the potential component life.

A reduction of the applied loading or an increase of component performance wouldresult in an increase of the safety margin.

Figure 1-2 illustrates the combined influence of the above considerations with a possibleevolution of safety requirements [58]. Some additional considerations on safetyrequirements are included in Section 4.

10

As shown on these figures, life extension is not only obtainable through repair orreplacement of components, but also through a better use and a better evaluation of thereal evolution of component performance, for example a better prediction of materialproperties evolution, a better evaluation of existing defects and mechanical behaviour ofreal defects, or a better knowledge of operating conditions.

A lot of studies may consequently be classified as contributing to ageing and plant lifemanagement, and bibliography syntheses referred to in Section 2 and in Appendices 1and 6 provide more information on this topic.

)LJXUH�� Illustration of combined evolution of component margins and safetyrequirements.

)LJXUH�� Illustration of possible component margin evolution during service.

Parameter

Time

Requiredmargin

Acceptable life

Structure resistance

Applied loading

Safetymargin

Time

Rough evaluation ofcomponent performance

Safety requirements

Acceptable life

Improvement ofcomponent performance

11

This report is more specifically dedicated to the safe management of ageing and will befocussed on:

− the presentation of the work done under international auspices. This aspect is coveredin Section 2, supplemented by Appendix 1,

− the presentation of the practices of ageing management in three European countries:Belgium, France and Spain. This aspect is covered in Section 4 for regulatoryaspects and in Section 5 for ageing management strategies, supplemented byAppendices 2 and 3,

− the comparison of theses practices with international recommendations. This iscovered in Section 6, supplemented by Appendices 4 and 5.

For a better understanding of sections 4 to 6, Section 3 includes a summary of thevarious degradation phenomena and their governing parameters, the methods ofidentifying ageing in service, and the possible mitigation methods of ageing effects.

The references specifically used or referred to in the main text of this report are listed inSection 8. An extended bibliography is given in Appendix 6.

Conclusions and recommendations are provided in Section 7.

�� 25*$1,6$7,216�,192/9('�,1�7+(�352-(&7Three countries are involved in the project: Belgium, France and Spain, represented bythe following organisations:

− Belgatom is a subsidiary of Tractebel and Belgonucléaire, acting on activitiesincluding, among others, application of industrial codes, surveillance, managementof time life and dismantling, and repair / replacement.

− EDF (Electricité de France), is the French national electricity utility, in charge ofbuilding and operating electricity generating plants. Its activities also includeengineering, life management, R & D, surveillance and non-destructiveexaminations, and repair / replacement.

− FRAMATOME, ensuring the general management of the project, is the French baseddesigner and manufacturer of nuclear reactors. The Group offers products, servicesand maintenance related to nuclear realisations, fuel fabrication, connectors andindustrial equipment.

− IPSN (Institute for Nuclear Safety and Protection) is the technical support of theFrench Safety Authority, in charge of preparing safety analyses for the NuclearInstallations Safety Directorate (DSIN), or reports discussed by the standing group ofexperts, to prepare DSIN decisions.

− UNESA is the Spanish Electricity Association, commissioned by its members tocarry out activities on the electricity business. In the nuclear area, it coordinatesaspects relating to nuclear plant safety and radiological protection, regulation andoperation, and technology engineering, as well as R & D.

13

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The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958under the name of OEEC European Nuclear Agency. It received its present designationon 20th April 1972, when Japan became its first non-European full Member. NEAmembership today consist of all OECD Member countries, except New Zealand andPoland. The Commission of the European Communities takes part in the work of theAgency (for more information, visit the NEA web site: http://www.nea.fr).

The primary objective of the NEA is to promote co-operation among the governmentsof its participating countries in furthering the development of nuclear power as a safe,environmentally acceptable and economic energy source.

The NEA works in close collaboration with the International Atomic Energy Agencyin Vienna, with which it has concluded a Co-operation Agreement, as well as withother international organisations in the nuclear field.

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Four committees develop actions relating with ageing or life management (fig. 2.1):

• the Committee on Nuclear Regulatory Activities (CNRA),

• the Committee on the Safety of Nuclear Installations (CSNI),

• the Nuclear Development Committee (NDC),

• the Nuclear Science Committee (NSC).

2.1.1.2.1. Committee on Nuclear Regulatory Activities

The Committee on Nuclear Regulatory Activities (CNRA) of the OECD NuclearEnergy Agency (NEA) is an international committee made up primarily of seniornuclear regulators. It was set up in 1989 as a forum for the exchange of informationand experience among regulatory organisations and for the review of developmentswhich could affect regulatory requirements.

The Committee focuses primarily on power reactors and other nuclear installationscurrently being built and operated. It also may consider the regulatory implications ofnew designs of power reactors and other types of nuclear installations.

The Committee is responsible for the programme of the NEA, concerning theregulation, licensing and inspection of nuclear installations.

14

2.1.1.2.2. Committee on the Safety of Nuclear Installations

The NEA Committee on the Safety of Nuclear Installations (CSNI) in an internationalcommittee made up of scientists and engineers. It was set up in 1973 to develop andco-ordinate the activities of the Nuclear Energy Agency concerning the technicalaspects of the design, construction and operation of nuclear installations insofar as theyaffect the safety of such installations. The Committee’s purpose is to fosterinternational co-operation in nuclear safety amongst the OECD Member countries.

CSNI constitutes a forum for the exchange of technical information and forcollaboration between organisations which can contribute, from their respectivebackgrounds in research, development, engineering or regulation, to these activitiesand to the definition of its programme of work. It also reviews the state of knowledgeon selected topics of nuclear safety technology and safety assessment, includingoperating experience. It initiates and conducts programmes identified by these reviewsand assessments in order to overcome discrepancies, develop improvements and reach

)LJXUH��Organisation chart of the NEA

15

international consensus in different projects and International Standard Problems, andassists in the feedback of the results to participating organisations.

The greater part of CSNI’s current programme of work is concerned with safetytechnology of water reactors. The principal areas covered are operating experience andthe human factors, reactor coolant system behaviour, various aspects of reactorcomponent integrity, the phenomenology of radioactive releases in reactors accidentsand their confinement, containment performance, risk assessment and severe accidents.The Committee also studies the safety of the fuel cycle and operates an internationalmechanism for exchanging reports on nuclear power plant incidents.

CSNI works through five Principal Working groups, (PWG’s) dealing with OperatingExperience and Human Factors (PWG-1), Coolant System Behaviour (PWG-2),Integrity of Structures and Components (PWG-3), Accidental Radioactivity Release(PWG-4) and Risk Assessment (PWG-5) (figure 2.2).

)LJXUH��Previous structure of CSNI.

Since 1996 PWG-3 has had 3 sub-groups, dealing with the integrity of metal structuresand components, the ageing of concrete structures, and the seismic behaviour ofstructures. Typical activities are the organisation of Workshops and Specialists

C hairma n :K. KathollS ecretary :

L. C arlsson

C hai rm an :M. RéocreuxSecreta ry :

M . H rehor

Chai rman :H. Sc hu lzSecreta ry :

A. M i ller

Chairm an :D. De B oekS ecretary :

J . Royen

NEA C o-operation w ithC EEC and N IS

S upport G roup on V VER-440/213 Bubbler CondeserConta inment Research

Sup port Gro up o nVVER TH Cod eValid ation Matrix

Sup port G ro up o nVV ER -1 000 Larg e S ca leTes t Faci li tie s (PSB )

W ork ing Group onFuel Cycle Sa fety

Chairman :J .A. MurphySecretary :

D . Ka ufer

Task Group onS afety

Research inMember

Countries

Expanded taskForce on Human

Factors

ThermalhydraulicApplication task

Group

Fuel IncidentNotificat ion andAnalysis Sys.

(FINAS)

Incident ReportingSystem (IRS)

Com puter-BasedSafety Sys tems

Degraded CoreCooling Task

Group

Task Force onFuel Safety

criteria

ICDE Project-Common Cause

Failure Data

Task Group onContainmentA spects of

S evere A ccidentManagement

Task Group onS evere A ccidentPhenomena in

the Containment

Task Group onFission ProductPhenomena in

the PrimaryCircuit and theContainment

Task Group on FireR isk Assessment

Task Group onHuman R eliabi lity-

Errors ofCommission

Task Group onSoftware

Reliability

Topical OpinionPapers, other PSAstudies and Data

Base Support

PW G-1O perating

Experience &Human F actors

PW G-2 CoolantSystem

Beha viour

PW G-3 In tegrityof Com ponents

and Structures

P W G-4Confinement of

AccidentalR adioactive

R eleases

PW G-5

Risk Assessment

S ESAR /FAP

CO M MITTEE O N TH E SA FETY O F NU C LE A R IN STA L LA TIO N (CSNI)Chairman : M. Livolant – S ecretary : G.M. F resc ura

C hai rman :M. Kanamorf

Sec ret ary :B . Kaufer

Task Group onSafety

Research inMem ber

Countries

Fuel Inc identNotification andA nalys is Sys.

(FINAS)

Ch airm an : T. K in gS ecretary : J . R oyen

16

Meetings, the preparation of State of the Art Reports, organisation of round robinexercises such as PISC in NDE or FALSIRE in fracture mechanics.

Ageing of active components (such as pumps, valves, diesels ... ) is covered by PWG-1and ageing of passive components and structures by PWG-3, but interestinginformation about ageing trend of components can be also gained in the studies ofPWG-5 on reliability data base.

At the beginning of 2000, restructuring of the five Principal Working Groups into fourWorking Groups has been engaged. This would be accomplished by merging theformer PWG 2 and PWG 4 into a single Working Group (fig. 2.3)..

)LJXUH��Revised structure of CSNI

2.1.1.2.3. Nuclear Development Committee

The Committee for Technical and Economic Studies on Nuclear Energy and the FuelCycle, known as the Nuclear Development Committee (NDC) has the following scope:

• Assessment of the potential future contribution of nuclear energy to overall energydemand ;

• Assessment of demand and supply for the different phases of the nuclear fuel cycle;

COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATION (CSNI)Chairman : M. Livolant – Secretary : G.M. Frescura – CSNI Bureau : A. Thadani, A. Alonso

Working Groups CSNI Programme Group Special Expert Groups

RiskAssessment

Human andOrganisational

Factors

Fuel SafetyMargins

Chairman :Secretary :

Analysis andManagementof Accidents

Integrity ofComponents

Structure

OperatingExperience






HumanReliabilityErrors of

Commission

Design Basis Fuel Cycle SafetyIntegrity of MetalComponents and

Structures

SoftwareReliability

Severe Accidents

Fission ProductBehaviour

ConcreteStructures Ageing

SeismicBehaviour ofStructures

DATA BasesIncident ReportingSystem (IRS)ICDE Project-CommonCause Failure DataComputer-Based SafetySystemsFuel incident & ReportingSystem (FINAS)

NEA Co-operation withCEEC and NIS

Support Group on VVER-440/213 Bubbler CondeserContainment Research

Support Group onVVER TH CodeValidation Matrix

Support Group onVVER-1000 Large ScaleTest Facilities (PSB)

RASPLAY

HALDENREACTORPROJECT

SandiaLower Head

FailureProject

17

• Review of the technical and economic characteristics of nuclear energy growth andof the nuclear fuel cycle ;

• Evaluation of the technical and economic consequences of the various strategies forthe nuclear fuel cycle.

The NDC set up an Expert Group on Nuclear Power Plant Life Management (PLIM) in1991. The main scope and objectives of the PLIM Expert Group are to assist decisionmakers at the Member state’s government level in evaluating the economics andpolitics of plant life extension by providing a published report in which the followingtypes of key issues are addressed in broad terms:

• Rationale for plant life extension in different NEA Member countries (economics,financial, reliability, availability, safety, etc ... ) ;

• Financial and economic concerns including intermediate results at incremental timeperiods of extension (5, 10, 15 and 20 years) ;

• Major technical concerns and constraints ;

• Institutional issues in and public acceptance on the plant life extension ;

• Need for and feasibility of creating data base which will encompass a wide rangeof data for plant life extension ;

• Precedents in other NEA Member countries and IAEA Member states.

It produced at 1997 a draft of an International Terminology on Plant Ageing (with thecooperation of IAEA and CEC). The final document entitled "Common Ageingterminology" was issued in July 1999.

In 1997 a group of experts was set up to collect and evaluate nuclear power plantrefurbishment cost data and experience accumulated over the last years in participatingcountries. This information may be useful to reactors operators faced with nuclearplant life cycle evaluations (issued 1999). New orientations for the PLIM Expert Groupare under discussion.

An international workshop was organised on 26 and 27 June 2000 to review theprospects for PLIM and the motivation for Research and Development to increaseconfidence investment to extend plant life. It was hosted by the United StatesDepartment of Energy (US DOE) in Washington, D.C, USA. A selection of paperspresented is included in Appendix 6.

2.1.1.2.4. Nuclear Science Committee

This Committee has some relevant activities on neutron dosimetry benchmarks andmaterial degradation modeling.

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The following documents with a link to ageing are summarised: see third column oftables below and summaries in Appendix 1.

18

2.1.1.3.1. CNRA

Summary withNo Reference Title conclusions and

recommendations(Y = yes; N = no).

Correspondingsheet in Append.1

[1] NEA Future nuclear regulatory challenges - November Y66 98 10 P 1998 A.1

ISBN 92.64.16106-6[2] NEA/CNRA/R(99)1 Regulatory Aspects of Ageing Reactors - March Y

1999 A.2

2.1.1.3.2. CSNI

3*:�


recommendations(Y yes; N no).Corresponding

sheet in Append.1[3] NEA/CSNI/R(95)1 State of the art on key fracture mechanics aspects of

integrity assessment, 1996 N(also referenced as: OCDE/GD(96)6)

[4] NEA/CSNI/R(95)4 Report on round robin activities on the calculationof crack opening behaviour and leak rates for small N

bore piping components, 1995(also referenced as OCDE/GD(95)90)

[5] NEA/CSNI/R(95)6 Workshop on Reactor Coolant System Leakage andFailure Probabilities (1992: Köln, Germany), 1995 N

(also referenced as OCDE/GD(95)91[6] NEA/CSNI/R(95)17 International Workshop on Aged and Decommiss-

ioned Material Collection and Testing for Structural YIntegrity Purposes (1995 : Mol, Belgiurn), 1996 A.3

(also referenced as: OCDE/GD(96)10)[7] NEA/CSNI/R(95)18 Leak before break in reactor piping and vessels

specialists meeting (1995: Lyon, France), 1996, NVols 1-3 (also referenced as: OCDE/GD(96)11

[8] NEA/CSNI/R(95)19 Report of the task group reviewing national andinternational activities in the area of ageing of Ynuclear power plant concrete structures, 1996 A. 4

(also referenced as: OCDE/GD(96)31)

[9] NEA/CSNI/R(96)1 FALSIRE : phase 2 : CSNI project for FractureAnalyses of Large-Scale International Reference N

Experiments, 1996(also referenced as: OCDE/GD(96)187)

19



sheet in Append.1[10] NEA/CSNI/R(96)4 Probabilistic structure integrity analysis and its

relationship to deterministic analysis (1996: NStockholm, Sweden), 1996

(also referenced as: OCDE/GD(96)124)[11] NEA/CSNI/R(96)10 Seismic shear wall ISP: NUPEC’s seismic ultimate

dynamic response test: comparison report, 1996 N(also referenced as: OCDE/GD(96)188)

[12] NEA/CSNI/R(96)11 Report of the task group on the seismic behaviourof structures : status report, 1997 Y

(also referenced as: OCDE/GD(96)189) A.5

[13] NEA/CSNI/R(97)1 NDE Techniques capability demonstration andinspection qualification : proceedings of the Joint N

EC, OCDE IAEA Specialists Meeting (1997:Petten, The Netherlands), 1997

(also referenced as: EUR 17354 EN)[14] NEA/CSNI/R(97)8 Fatigue crack growth benchmark N[15] NEA/CSNI/R(97)9 Joint WANO/OECD-NEA Workshop: Prestress Y

loss in NPP containments (1997: Poitiers, France), (conclusions in1997 (also referenced as: OCDE/GD(97)225) ref. [22])

A.6[16] NEA/CSNI/R(97)28 Development priorities for NDE of concrete N

structures in nuclear plants (NEA Workshop, (conclusions inRisley, United Kingdom, Nov. 97) 1998 ref. [18])

[17] NEA/CSNI/R(98)5 Status report on seismic re-evaluation - Nov. 1998 YA.7

[18] NEA/CSNI/R(98)6 Development priorities for non-destructive Yexamination of concrete structures in nuclear plant,

Nov. 1998A.8

[19] NEA/CSNI/R(98)7 Survey of organic components in nuclear power Yplants, 1998 A.9

[20] NEA/CSNI/R(98)8 Experience with Thermal Fatigue in LWR Piping YCaused by Mixing and Stratification - December

1998A.10

[21] NEA/CSNI/R(98)9 PISC III: Final report, 1998 N[22] NEA/CSNI/R(99)1 Finite Element analysis of degraded concrete Y

structure - Workshop Proceedings - BNL, NY,USA, 29-30 Oct. 1998 - Sept 99

A.11

[23] NEA/CSNI/R(99)11 NPP Containment Prestress loss. Summary YStatement - Sept. 99 A.12

[24] OECD/PWG 3 Relation of ageing and seismic engineering - Draft - YJune 1999 A.13

[25] CSNI - PWG 3 Plant Ageing Management - Providing a technical YTechnical Position basis for long-term operation of light water reactors A.14

Document Draft - May 1999[26] Transaction of the Activities of the OECD Nuclear Energy Agency in N

SMIRT15- Seoul, the area of concrete containment ageing - A.

20



sheet in Append.1Korea MILLER and L. SMITH

August 15-20,1999

3:*��

No Reference Title Summary(Y = yes; N = no).

Correspondingsheet in Append.1

[27] NEA/CSNI/R(95)9 Evidence of Ageing Effects on Certain Safety -Related Components Y

Volume 1 : Summary and Analysis (68 pages) A.15Volume 2: Contributions - RESTRICTED

2A: France, Sweden - 2B: Finl - 2C: Japan, US, UKSeptembre 1995

[28] NEA/CSNI/R(97)23 Operating and Maintenance Experience withComputer-based Systems in NPPs - Septembre

1998 (53 pages)Y

A16

3:*��


recommendations(Y = yes; N = no)

[29] NEA/CSNI/R(98)10 Reliability Data Collection - WorkshopProceedings - Budapest, Hungary (21-23 April

1998) - March 1999 (260 pages)N

21

6(6$5�)$3��6HQLRU�*URXS�RI�([SHUWV�RQ�1XFOHDU�6DIHW\�5HVHDUFK�)DFLOLWLHV�DQG3URJUDPPHV�


recommendations(Y = yes; N = no)

[34] SESAR-FAP Major Facilities and Programmes at risk - Draft,18 August 1999 N

2.1.1.3.3. NDC


recommendations(Y yes; N no)Corresponding

sheet in Append.1[30] NEA/SEN/NDC(97)11 PLIM Workshop - 6th Meeting of the Expert Group

Rev. 1 on Nuclear Power Plant Life Management - June Y1997 (229 pages) A.17

[31] Common Ageing Terminology - July 1999 (Joint Nwork of NEA, CEC and IAEA; in five language)

[32] NEA/NDC/DOC(99)1 Refurbishment costs of Nuclear Power plants - YJanuary 1999 (76 pages) - RESTRICTED A.18

[33] Policy and Effective Management of Nuclear PowerPlant Life Management - First draft version 5 N

April 15,1999

�� 6\QWKHVLV�RI�2(&'�1($�GRFXPHQWV�RQ�DJHLQJ

CNRA and CNSI consider ageing issues as important for the future. A distinction ismade between the physical ageing of components and structures and other ageingconcerns : ageing of analytical techniques and documentation, ageing of rules andstandards, ageing of technology, ageing of organisation, ageing of plant personnel. Thefirst domain is linked to the demonstration of the structural integrity of the componentsand structures throughout the lifetime of a plant. The second domain deals moreglobally with the management of change.

Managing the physical ageing of the plant components and structures is the highestpriority but no real concerns are expressed in regard of safety. It is considered that theageing of active components for functional aspects is correctly managed by the overallmaintenance and surveillance programmes. The management problem is mainly withpassive parts or passive components.

Needs for further research and development are indicated in the following areas:

Metallic components and structures:

• better understanding of the ageing phenomena affecting the pressure boundary ofthe primary systems (fatigue, thermal and irradiation embrittlement, thermal shock,corrosion, erosion and cracking, crack initiation and propagation under the variousenvironmental conditions prevailing),

22

• to test materials from decommissioned reactors to improve the knowledge ofirradiation effects and other ageing degradations and to gain more information ondefect distributions.

Concrete structures

Performance of the structures has been good, nevertheless developments are needed:

• techniques for in-service inspection of thick sections and inaccessible areas(determination of as-built structural details ; detection of flaws ; characterisationand quantification of flaws) and inspection techniques for anchorage and sensitiveparts,

• durability of remedial measures and repairs,

• loss of prestressing force in tension of post-tensioned concrete structures,

• validation of methods for the finite elements analysis of degraded structures.

Organic components

Research is needed to develop more realistic end of life criteria or forecast lifetime interms of the functionality of a component and condition monitoring methods.

Relating to the management of change, to provide a technical basis for long-termoperation of nuclear power plants it is necessary to :

• update the individual plant documentation to avoid gaps in knowledge,

• establish a system of information retrieval to bridge gaps between today’s andprevious design and manufacturing standards and safety rules,

• determine more precisely minimum functional requirements to allow qualificationof new equipment (standard ones when possible),

• research needs to ensure safety of replacement of ageing Instrumentation andControl are to be defined in order to supply as much as possible standards modules,

• ageing of organisation should be managed : specific skills shall be identified andtransferred when supplies disappear,

• taking into account the expected length of life time, several "generations" ofpersonnel shall successively insure that the knowledge of design criteria, thereasons of choice of solutions are always known : transfer of knowledge is animportant challenge,

• enforce clubs of users of similar technology internationally.

�� ,$($

�� *HQHUDO�SUHVHQWDWLRQ�RI�,$($

The IAEA is an autonomous intergovernmental organisation within UN family whichwas founded in 1957 in accordance with a decision of the General Assembly of theUnited Nations. Its Statutory mandate is "to accelerate and enlarge the contribution ofatomic energy to peace, health and prosperity throughout the world and to ensure, sofar as it is able, that assistance provided by it or at its request or under it supervision orcontrol is not used in such a way to further any military purpose". It is authorised tofoster research and development in the peaceful uses of nuclear energy and theexchange of scientific and technical information, to establish and administer safeguardsagainst the diversion of nuclear materials to military purposes that were intended for

23

use in civil nuclear programmes and to establish or adopt health and safety standardsregarding nuclear energy.

�� $FWLYLWLHV�LQ�WKH�ILHOG�RI�DJHLQJ�DQG�OLIH�PDQDJHPHQW

These activities are carried out in the Division of Nuclear Power for life managementand technological aspects and in Division of Nuclear Installation Safety for moresafety-related aspects.

2.1.2.2.1. Safety related aspects : Safety Standards and guidance

6DIHW\�6WDQGDUGV

General guidance on NPP activities relevant to the management of ageing(maintenance, testing, examination and inspection of Systems, Structures andComponents (SSC)) is given in the Safety Standards (NUSS) Code on the Safety ofNuclear Power Plants : Operation (No 50-C-0, Rev. 1) and associated Safety Guides onIn-service inspection (50-SG-02), Maintenance (50-SG-07, Rev. 1) and Surveillance(50-SG-08, Rev. 1). All these documents are being revised :

• 50-C-0 will become "Requirements for the Safety of Nuclear Power Plants :Operation"

• 50-SG-02, 07 and 08 will be combined in "Maintenance, Testing, Surveillance andIn-Service Inspection of Nuclear Power Plants".

Other publications deal with a comprehensive safety review of NPPs which includesnon-physical ageing, resulting in SSCs, procedures, documentation, etc. becoming out-of date in comparison with current standards, methods and technology :

• Safety Reports Series No 12 : Evaluation of Safety of Operating NPPs Built toEarlier Standards,• Safety Guide 50-SG-012 : Periodic Safety Review of Operational NPPs.

Moreover it is intended to put in each standard a paragraph on ageing (if necessary).

*XLGDQFH

The IAEA initiated activities on safety aspects of NPP ageing in 1985 to increaseawareness of the emerging safety issue relating to physical ageing of plants systems,structures and components (SSCs) and in 1989 a systematic project aimed at assistingMember States in understanding ageing of SSCs important to safety and in effectiveageing management of these SSCs in order to ensure their integrity and functionalcapability throughout their service life. This project integrates information on theevaluation and management of safety aspects of NPP ageing generated by MemberStates into a common knowledge base, and derives guidance and assists Member Statesin the application of this guidance. Main results of the projects are documented in thelist of selected IAEA documents of paragraph 2.1.2.3. They fall into four groups.

$ZDUHQHVV.

Following up on the first International Conference on Safety Aspects of Ageing andMaintenance of Nuclear Power Plants which was organised by the IAEA in 1987,increased awareness of physical ageing of SSCs and its potential safety impact wasachieved by the development and wide dissemination in 1990 of an IAEA-TECDOCon Safety Aspect of Nuclear Power Plant Ageing [35]. While in the 1980s most peoplebelieved that classical maintenance programmes were adequate for dealing with theageing of nuclear plants, in the 1990s the need for ageing and life management ofNPPs became widely recognised.

24

3URJUDPPDWLF�*XLGHOLQHV��The following programmatic guidance reports have beendeveloped using experience of Member States.

'DWD� &ROOHFWLRQ� DQG� 5HFRUG� .HHSLQJ� IRU� WKH� 0DQDJHPHQW� RI� 1XFOHDU� 3RZHU� 3ODQW$JHLQJ [36]. provides information on the baseline, operating and maintenance dataneeded and a system for data collection and record keeping.

0HWKRGRORJ\� IRU� WKH� 0DQDJHPHQW� RI� $JHLQJ� RI� 1XFOHDU� 3RZHU� 3ODQW� &RPSRQHQWV,PSRUWDQW� WR�6DIHW\� >��@�gives guidance on screening SSCs to make effective use of

)LJXUH��,$($�2UJDQLVDWLRQDO�FKDUW

25

limited resources and on performing ageing management studies to identify or developeffective ageing management actions for the selected components.

,PSOHPHQWDWLRQ� DQG� 5HYLHZ� RI� 1XFOHDU� 3RZHU� 3ODQW� $JHLQJ� 0DQDJHPHQW3URJUDPPHV�>��@ provides information on the systematic ageing management processand an organisational model for its implementation.

(TXLSPHQW�4XDOLILFDWLRQ�LQ�2SHUDWLRQDO�1XFOHDU�3RZHU�3ODQWV�>��@ documents currentmethods and practices relating to upgrading and preserving equipment qualification inoperational NPPs and reviewing the effectiveness of plant equipment qualificationprogrammes.

&RPSRQHQW� VSHFLILF� JXLGHOLQHV� TECDOC documents on Assessment andManagement ageing of Major NPP Components Important to Safety

The guidance of Ref. [37] has been used to implement Co-ordinated Research Projects(CRPs) on management of ageing of concrete containment buildings and in-containment instrumentation and control cables (see §2.1.2.2.3), and to developcomprehensive technical documents on Assessment and Management of Ageing ofMajor Nuclear Power Plant Components Important to Safety.

The objectives of the report series are :

• to help ensure the functional capability of selected NPP components important tosafety by documenting current practices on ageing assessment and management ofthese components,

• to provide a common technical basis for a dialogue between NPP operators andregulators.

These reports are component specific ; they are focused on safety perspective ;economics and therefore life management aspects are not addressed.

The structure is common to all the reports :

• component description• component design basis• potential ageing mechanisms• operating guidelines• inspection and monitoring requirements, techniques and practices• methods for the assessment of degradation• maintenance methods• systematic component specific ageing management programme (AMP).

The LWR components selected are :

• Steam Generators [40]• Concrete Containment Buildings [41]• PWR Reactor Pressure Vessels [42]• PWR RPV Internals [43]• BWR RPV Internals (in preparation)• Metal components of BWR Containment Systems (in preparation)• PWR Primary Piping (near completion)• Instrumentation and Control Cables (end 1999 ; linked to CRP § 2.3)

A technical document presenting good practices in the Implementation ofConfiguration Management in NPPs is in preparation (management of the ageing ofdocumentation).

26

$JHLQJ� PDQDJHPHQW� UHYLHZ� JXLGHOLQHV�� $0$7� *XLGHOLQHV� >��@� is a referencedocument for IAEA Ageing Management Assessment Teams (AMAT) and for utilityself-assessments ; these reviews can be programmatic or problem oriented.

The focus of the project work has progressively shifted from developing awareness, topreparing first programmatic, and then component specific guidelines. In the future,the focus will be on providing services to assist Member States in the application of theguidelines. A reduced effort will be maintained to facilitate information exchangethrough the preparation of additional guidelines and the updating of existingguidelines.

*XLGDQFH�LQ�SUHSDUDWLRQ

• Methodology for SSC selection and safety based prioritisation of AgeingManagement actions• Guidance on minimising premature ageing• Guidance on optimisation of NPP maintenance in support of ageing management.

2.1.2.2.2. Life Management and Technological Aspects

In the Division of Nuclear Power, these activities are co-ordinated by the InternationalWorking Group on Life Management of Nuclear Power Plants (IWG-LMNPP).

This group set up twenty years ago is now composed of representatives from 27countries(*) and 2 international organisations (OECD-NEA and EC).

The IWG-LMNPP covers the following aspects :

• Design• Materials• Fabrication• Monitoring, Testing, Inspection and data bases of their results• Degradation mechanisms, their significance and mitigation• Assessment and means of plant life management• Strategic, economic and administrative aspects of NPP life management and

decommissioning.

Its topical priorities are :

• Reactor Pressure Vessel (RPV) integrity• Steam Generator life Management• Primary Circuit Operation and Integrity• Reactor Internals Operation and Integrity• Secondary Circuit• Containment/civil structures• Cables• Other items of importance including economic aspects of life management.

In recent years the IWG-LMNPP has organised several specialists meeting :

• Steam generator repair and replacement, practices and lessons learned (1996)• Methodology of pressurised thermal shock evaluation (1997)• Irradiation effects and mitigation (1997)• NPP condition Monitoring and Maintenance (1998)• NPP life management policies and strategies (1998)• Behaviour of core internals (1998).

(*) from which all western and eastern European countries.

27

and in co-ordination with other international organisations :

• NDE techniques capability demonstration and inspection qualification (1997 ; jointEC/OECD-NEA/IAEA meeting)

• Non destructive methods for monitoring degradation (1999; joint EC/IAEAmeeting)

The next meetings planned are :

• Irradiation embrittlement and mitigation (1999)• Erosion/corrosion of NPP components (1999)

Besides these meetings and the regular meetings of the IWG-LMNPP, Research Co-ordination meetings are organised for the participants to Co-ordinated ResearchProjects (CRP):

• CRP "Management of ageing reactor pressure vessel primary nozzle" (This CRP isended)

• CRP "Assuring structural integrity of reactor pressure vessel" (completed in 1999).

IAEA has also taken the initiative and proposed the establishment of a database onageing management and life extension of NPP key components important to safety andproductivity (reactor pressure vessel / primary piping / containment / other keycomponents) ; at the moment only the international RPV material surveillance databaseis well developed.

2.1.2.2.3. Research contract programme

Pilot studies were initiated at IAEA in 1989 related to the evaluation and managementof the safety aspects of nuclear power plant ageing. The purpose of the studies was tofacilitate the exchange of information and collaboration among internationalorganisations engaged in ageing management and evaluation projects. This led to thedevelopment of an overall strategy for IAEA-coordinated ageing management studiesinvolving four components (see ref. 55) :

(1) reactor pressure vessel primary nozzle,(2) motor-operated isolating valve,(3) concrete containment building,(4) in-containment instrumentation and control cables.

Studies on components (1) and (2) are sponsored by the Nuclear Power Division withsponsorship of components (3) and (4) by the Nuclear Installation Safety Division.These components were selected on the basis that they represent different safetyfunctions and material, as well as susceptibility to different ageing mechanisms.

The general objectives of the ageing studies for each of these components were toidentify dominant ageing mechanisms and to identify or develop an effective strategyfor managing ageing effects caused by the identified mechanisms. Results of thesestudies were expected to have application in :

(1) monitoring the degradation and in preventative maintenance of the selectedcomponents (including the development of criteria for designs of the type andtiming of preventative maintenance actions),

(2) predictions of component performance and remaining service life under allexpected service conditions,

(3) future designs and material selections,

(4) amendments to applicable codes, standards and regulatory requirements.

28

The resulting pilot studies were implemented in two phases. Phase I pilot studies (i.e.,interim ageing studies) were completed through Technical Committee Meeting. UnderPhase I dominant ageing mechanisms were identified as well as a strategy formanaging ageing.

Phase II activities (comprehensive ageing studies) have been implemented throughIAEA Co-ordinated Research Projects (CRP) that take place over a three- to four-yearperiod.

The status of these CRP is as follows :

(1) RPV primary nozzle : ended in 1999; report pending

(2) Motor-operated isolation valve : phase 4 stopped in 1998 with only partialachievement - Report pending

(3) Concrete containment building : completed in 1996

(4) In containment instrumentation and control cables : end foreseen by mid 2000.

In 1999 some new CRPs have been launched (Department of Nuclear Energy). Theyare on :

• Ageing of materials in spent fuel storage facilities.• Mechanism of Nickel effect in radiation embrittlement of RPV steels• Surveillance programmes results application to RPV integrity assessments

�� /LVW�RI�VHOHFWHG�,$($�GRFXPHQWV�UHODWLQJ�WR�DJHLQJ�DQG�OLIHPDQDJHPHQW

Reference TitleSummary withconclusions and

recommendations(Y = yes ; N = no)

Correspondingsheet in Append. 1

[35] TECDOC-540 (1990) Safety Aspects of Nuclear Power Plant Ageing YB.1

[36] Safety Series No 50-P.3 (1991)

Data collection and Record keeping for themanagement of Nuclear Power Plant Ageing - A

Safety Practice

YB.2

[37] Technical ReportsSeries No 338

1992

Methodology for the Management of Ageing ofNuclear Power Plant Components Important to

Safety

YB.3

[38] Safety Series No 151999

Implementation and Review of a Nuclear PowerPlant Ageing Management Programme

YB.4

[39] Safety Reports SeriesNo 3 - 1998

Equipment Qualification in Operational NuclearPower Plants

YB.5

[40] TECDOC-981November 1997

Assessment and management of ageing of majornuclear power plant components important to

safety : Steam Generators

YB.6

[41] TECDOC-1025June 1998

Assessment and Management of Ageing ofMajor Nuclear Power Plants Components

Important to Safety : Concrete ContainmentBuildings

YB.7

[42] TECDOC-1120October 1999

Assessment and management of ageing of majornuclear power plant components important to

YB.8

29




safety : PWR Pressure Vessels[43] TECDOC-1119

October 1999Assessment and management of ageing of major

nuclear power plant components important tosafety : PWR Vessel Internals

YB.9

[44] Services Series No 4March 1999

AMAT Guidelines. Reference document for theIAEA ageing management assessment teams

N

[45] TECDOC-932March 1997

Final Report : Pilot Studies on Management ofAgeing of Instrumentation and Control CablesResults of a co-ordinated research programme

1993-1995

N

[46] IAEA/NSNI - Reporton the IAEA

ResearchCoordination meeting

- 8-12 June 1998,EDF, Bordeaux,

France

Co-ordinated Research Programme (CRP) onManagement of Ageing of In-Containment

Instrumentation and Control Cables - LIMITEDDISTRIBUTION

YB.10

[47] Safety Reports SeriesNo 121998

Evaluation of the Safety of Operating NuclearPower Plant Built to Earlier Standards - A

common Basis for Judgement

YB.11

[48] Safety SeriesNo 50-C-0 (Rev. 1)

1988

Code on the Safety of Nuclear Power Plants :Operation (In course of revision ; draft April

1999)

YB.12

[49] Safety SeriesNo 50-SG-09

1994

Management of Nuclear Power Plants for SafeOperation : A safety Guide

YB.13

[50] Safety SeriesNo 50-SG-012

1994

Periodic Safety of Operational Nuclear PowerPlants : A Safety Guide

YB.14

[51] Transactions of the15th International

Conference onStructural Mechanics

in ReactorTechnology(SMIRT15)

Seoul, Korea (August15-20, 1999)

Systematic Ageing Management Process : a keyelement for Long Term Safety, Reliability and

Economy of Nuclear Power Plants - J.Pachner/IAEA

N

[52] Internat. Conferenceon the Nuclear Power

OptionVienna - 5-9 Sep. 94

Aspects of Plant Life Assurance and Plant LifeManagement.

L.M. Davies, A.D. Boothroyd and L. IankoPaper IAEA-CN-59/40

N

[53] TECDOC-1084May 1999

Review of Selected Cost Drivers for Decisionson Continued Operation of Older Nuclear

Reactors - Safety Upgrades, Lifetime Extension,

YB.15

30




Decommissioning[54] IWG-LMNPP-94/6

1994International Database on Ageing Management

and Life Extension - Database SpecificationN

[55] TECDOC-670October 1992

Pilot studies on Management of Ageing ofNuclear Power Plant Components - Results of

Phase I

YB.16

[56] INSAG-14November 1999

Safe management of the Operating Lifetimes ofNuclear Power Plants

YB.17

�� 6\QWKHVLV�RI�WKH�,$($�'RFXPHQWV

The IAEA has developed a set of useful guidance for the systematic ageingmanagement process for physical or material ageing ([36], [37], [38], [44]).

Non-physical ageing is to be managed by periodic safety reviews ([47], [50]) and apermanent attention to these problems.

Moreover several IAEA NUSS codes and Guides may be of help for the managementof ageing ([39], [48], [49]) as well as comprehensive technical documents (TECDOC)focused on ageing ([35], [40], [41], [42], [45], [53], [55]).

As relates to the difficulties encountered to manage the physical ageing, the opinionsfound in the IAEA documents broadly support those of OECD-NEA. For example it isindicated that the ageing of active and relatively short-lived components can beadequately managed by existing operations, maintenance and qualificationprogrammes ([51]).

Critical components or structures are the same than those identified by NEA.

�� :$12

�� *HQHUDO�SUHVHQWDWLRQ�RI�:$12

The World Association of Nuclear Operators (WANO) unites all nuclear electricityoperators in the world. It facilitates the exchange of operating experience, so that itsmembers can work together to achieve the highest possible standards of safety andreliability in operating their nuclear power plants.

The establishment of WANO came from the accident at the Chernobyl nuclear powerplant in 1986. Beyond the immediate effects of the accident, there were far reachingrepercussions for the nuclear power industry as a whole. It caused nuclear operators toreassess the issue of safety and made them aware of the need for internationalcooperation.

Through WANO all nuclear power plant operators can communicate and exchangeinformation with one another within a culture of cooperation and openness.

WANO’s mission

31

WANO’s mission is to maximise the safety and reliability of the operation of nuclearpower plants by exchanging information and encouraging communication, comparisonand emulation amongst its members.

Experience shows that many accidents could have been prevented if lessons had beenlearned from previous incidents. The basic principle underlying WANO is that it isbetter to learn from someone else’s mistakes than from your own, and that it is better tohave the benefit of someone else’s good ideas rather than do all the work yourself withperhaps no better results.

WANO’s organisation

Membership of WANO is through one or more of its four regional centres, Atlanta,Moscow, Paris and Tokyo, and is determined by geographical location or reactor type.There is also a coordinating centre in London. The regional centres carry out theWANO Governing Board’s decisions and organise WANO’s programmes. Much oftheir work involves collecting, screening and analysing operating information beforesending it to members and ensuring WANO programmes meet member needs.

A central Governing Board provides the overall direction of WANO and establishesWANO policies.

WANO’s programmes

The WANO’s mission is implemented through 5 programmes which are:

• 7KH�2SHUDWLQJ�([SHULHQFH�,QIRUPDWLRQ�([FKDQJH�3URJUDPPH

This programme enables all members to learn from the operating experience of otherplants. In particular, the programme alerts members to events that have occurred atother plants and enables members to take action to prevent similar events fromhappening at their own plants. The programme is sometimes known as EventReporting.

• 7KH�2SHUDWRU�WR�2SHUDWRU�([FKDQJH�3URJUDPPH

This programme enables members to directly share plant operating experience andideas for improvement through face to face communication and by using the electronicmessaging system, NUCLEAR NETWORK®.

• 7KH�3HUIRUPDQFH�,QGLFDWRU�3URJUDPPH

This programme provides a means by which members can assess the performance oftheir plants objectively. There are ten performance indicators, which relate to nuclearplant safety and reliability, plants efficiency and personnel safety.

32

• 7KH�*RRG�3UDFWLFHV�3URJUDPPH

This programme enables members to learn from each other’s best practices and therebyimprove their own operational safety and reliability. A good practice is a technique,programme or process that has been proven particularly effective at improving safetyand reliability at one or more nuclear power plants.

• 7KH�3HHU�5HYLHZ�3URJUDPPH

This programme aims to help WANO members compare their operational performanceagainst best international practice through an in-depth, objective review of theiroperations by an independent team from outside their utility. The review, carried out atthe request of the plant, is conducted by an international review team consisting of stafffrom other members’ nuclear power plants.

�� /LIH�0DQDJHPHQW�DQG�7HFKQRORJLFDO�DVSHFWV

WANO itself does not produces or induces on the subject of ageing, analysis of theevents which are reported by the different member utilities and the Nuclear powerplants they are operating through the Operating Experience Information ExchangeProgramme. It provides a data bank on all reported events, which is accessible to allmember operators.

Since the beginning of 1998 a codification system has been set up for all the eventreports provided by the utilities and plants. It is already used for the events reportedsince that date, and the WANO Experience Feed Back is codifying all the eventsreported previously.

The quite detailed coding system includes the following main fields:

1. Category under which the event was reported (8 sub fields),2. Consequence of the event (10 sub fields),3. Malfunctioning, failed, affected, degraded systems (107 sub fields),4. Malfunctioning, failed, affected, degraded components (45 sub fields),5. Status of the reactor at the time the event occurred or was detected (11 sub fields),6. Activity that was being performed at the time the event occurred or was detected

(23 sub activities),7. Group of staff most involved in, or likely to learn from the event (31 sub fields),8. Direct cause : The failure, action, omission or condition which immediately

produced (or-led to) the event (55 sub fields),9 and 10. Root cause and causal factors (190 sub fields)

When there are a lot of sub fields they are in 2 or 3 sub levels.

Out of all these codes only a few of them can be considered as linked to ageing. Theycan be:

• In "Direct cause" :

− loosening... (0101)− corrosion, erosion .... (0102)− fatigue (0104)− break, rupture, crack.... (0106)− wear... (0108)

• In "Root cause" :

− historical design does not meet current requirements (2009)

33

− ageing of component (2302)− degraded sub component contributed to failure (2304)− component beyond expected lifetime (2306)− equipment erosion/corrosion (2308)− failed within expected lifetime (2309)

The events under these codes could be used as material for different kind of analysis.

�� ,17(51$7,21$/�'$7$%$6(6

�� ,$($

In the frame on the International Working Group on Life Management of NPPs (IWG--LMNPP) an International Database on NPP life management has been set-up.

The structure is composed of 5 databases (see fig 2.5)

1. International Database on reactor pressure vessel materials (IDPRVM)2. Database on pipe work on NPP3. Database on Steam Generators4. Database on NPP concrete containment5. Database on other NPP components

Only the first database (IDPRVM) is working at the moment, with the followingparticipation Belgium, Brazil, France, Hungary, Italy, Korea, Russia, Spain, Ukraine,USA.

It deals with

• material identification• ageing history• mechanical testing• visual data

)LJXUH��Structure of the International Database on Nuclear Power Plant Life Management

34

• references• evaluated data

For more information see "International Database on Ageing and Life Extension -Database specification" - IWG-LMNPP - 94/6, 1994 [54].

Database 2 (pipework) and 3 (Steam Generators) are in course of setting up. Theirstructure is indicated in the following figures 2.6 and 2.7.

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2.1.4.2.1. PWG 1

)LJXUH��Structure of the IAEA International Database on Pipework of Nuclear Power Plants

)LJXUH��Structure of the International Database on Steam Generators

35

,QWHUQDWLRQDO�&RPPRQ�&DXVH�)DLOXUH�'DWD�([FKDQJH��,&'(��'DWDEDVH : containsindependent failures as well as common cause failures for Pumps, Emergency DieselsGenerators (EDG), Motor Operated Valves (MOV), Safety Valves, Relief Valves andChecks Valves. Participants : Canada, Finland, France, Germany, Spain, Sweden,Switzerland, United Kingdom, USA.

,QWHUQDWLRQDO�'DWDEDVH�RQ�H[SHULHQFH�IHHGEDFN�RI�FRPSXWHULVHG�V\VWHPV.Participants: Japan, Finland, France, Germany, Hungary, Sweden, USA.

2.1.4.2.2. PWG 3

3LSLQJ�IDLOXUH�GDWDEDVH, in conjunction with EPRI (coordination), the WestinghouseOwners Group and the Swedish Nuclear Power Inspectorate - in course of setting-up(project completed by July 2000).

2.1.4.2.3. PWG 5

Since 1986 OECD-PWG 5 has arranged a series of international workshops on livingPSA in particular to facilitate international exchange on Reliability Data Collection.The last workshop, held on 21-23 April 1998 was focused on Reliability DataCollection (see ref. [29]).

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There are relatively few studies made under European Commission contracts on ageingmanagement, which is the reason why the present study was undertaken. Nevertheless,studies made on topics of interest to ageing management are numerous.

As shown in Section 1, life extension is not only obtainable through repair orreplacement of components, but also through a better use and a better evaluation of thereal evolution of component performance.

A lot of studies may consequently be classified as contributing to ageing and plant lifemanagement: for example: characterisation of materials dealt with in Working Groupon Codes and Standards (WGCS) Activity group Nr3, in-service inspection practicesand non-destructive examination techniques covered by WGCS activity group Nr1, andmechanical analysis tools dealt with in WGCS activity group Nr2.

Several institutions in Europe have capabilities to deal with several of the problemsposed by the ageing of structural components and their structural integrity assessment.These institutions and the Joint Research Center (JRC) have developed cooperativeprogrammes now organised in Networks [82]. They include representatives of utilities,engineering companies, R&D laboratories and Regulatory Bodies.

These networks are organised and managed like the PISC programme, the institute foradvanced materials of JRC playing the role of operating agent and manager of thesenetworks. Each network: ENIQ, AMES, NESC deals with a specific aspect of fitnessfor purpose of materials in structural components.

The intent of this Section is to summarise and highlight the main studies made underEuropean auspices, which may be of interest for ageing management.

36

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The strategic goal of the specific programme "Research and training programme in thefield of nuclear energy" [90] is to help exploit the full potential of nuclear energy, bymaking current technologies even safer and more economical and by exploring newconcepts. Ageing management clearly falls under the first of these approaches.

Under key action 2 "Nuclear fission", one of the first research objectives is related toimproved methods for understanding and managing the effect of ageing on equipmentand structures, for on-line monitoring (which may contribute to improve margins), andfor risk informed approaches to plant modernisation and inspection, which contributeto the demonstration and improvement of available safety margins.

Under topic 2.1, and particularly "Plant life extension and management", the 5-thFramework Programme expresses the following objectives: "develop a common basisfor the continued safe operation and prolonging the safe operational life-span ofexisting installations and to better methods for their inspection, maintenance andmanagement (both in terms of performance and occupational exposure)".

The following aspects are identified:

Integrity of equipment and structures: Understand, predict and properly manage theeffect of ageing, such as environmental assisted cracking, wear, fatigue, irradiationdamage, and dynamic loads on equipment and structures, and develop methods forestimating the influence of these effects on the safety margins under normal andabnormal operation.

Concerning embrittlement of components, the following projects may be mentioned[95] (see figure 2.7):

− FRAME will be focused on the improvement of the assessment of the RPVembrittlement through the development of a method allowing a direct measurementof the fracture toughness,

− RETROSPEC have the objective of improving the evaluation of the neutron doseswhere no or unreliable data from surveillance specimens are available,

− PISA intends to provide a better understanding of the role of phosphorus in theembrittlement process of RPV steels.

The following projects are focused on corrosion issues:

− INTERWELD deals with irradiation assisted cracking of austenitic steels used forreactor internals,

− CASTOC concerns environmentally assisted corrosion of low alloy steels understatic and cyclic conditions.

Prediction of structural safety margins are dealt with under ADIMEW project, whichhas the same objectives than FP-4 BIMET, and VOCALIST the intent of which is toimprove safety margins predictions.

FP-5 project WAHA/WHLOADS covers optimisation of operational conditions, inparticular dynamic loadings during operation.

On-line monitoring, inspection and maintenance: Development of innovative orimproved non-destructive methods and techniques for on-line monitoring and of risk-

37

informed approaches to inspection and maintenance. FP-5 projects related to this topicinclude:

− GRETE, which deals with NDT potential to support decision-making process forfailure prevention,

− IQNAR/SPICRACK devoted to the improvement of ultrasonic inspectionperformance for detection and sizing possibly present cracks in structuralcomponents,

− LIRES, which has the objective to develop electrodes likely to be used in harshoperational LWR conditions for corrosion potential monitoring.

Organisation and management of safety: Development of methods and tools, includingquantitative indicators, to assess the role of man, organisation and management inmaintaining and improving safety. Development of better methods and preventivemeasures that can be used for a risk based systems approach to plant modernisation,including PSA-techniques and applications. Development of man-machine interfaceand a common understanding of the qualification of advanced safety and controlsystems. Within FP-5, the following issues will be addressed:

− Computer-based systems to support I&C functions important to safety, within BE-SECBS project,

− The development of a safety justification framework for the refurbishment ofsystems important to safety, within CEMSIS project.

Networking: Following FP-4 INTACT concerted action, the VERSAFE project will befunded under FP-5 to cover plant life management and accident prevention andmitigation strategies.

The present study covers the whole above field, and intends to compare the existingnational approaches and to prepare recommendations on the complementary actions tobe undertaken in order to optimise the ageing management of nuclear power plants.

The above objectives intend to permit a better evaluation of plant performance (seefigure 1-2). Under topic 2.4 "Radiation protection", the programme also identifies riskassessment and management approach developments that can find technical and socialacceptance, governing safety requirements evolution (the second topic in figure 1-2).This aspect is also very important from ageing management point of view as far as theacceptable plant life is more and more identified with public acceptance criteria.

Consequently, the present study shall also deal with an improvement of the consistencyand presentation of ageing management concepts, capable of finding broad acceptance.

The 5-th Framework Programme also identifies support for research topics, in order toprovide tools and exchange of data needed for key actions implementation. Thoserelated to ageing concerns include facilities for degradation investigation, equipmentand material performance evaluation, data bases on safety assessment of majorcomponents, and supports to networks focusing on the major issues of the identifiedkey actions.

Training activities are also part of this programme. As far as ageing of expertise andprocedures are part of the global ageing evaluation, such activities shall also beconsidered in the recommendations to be prepared in the context of the present study.Exchanges, and special training courses fall under this scope. Maintaining the nuclearexpertise is a necessary condition for the current NPPs to continue to operate in a safeand competitive way.

38

Even if work is done on extreme severe accident mitigation, and if such action mayhave consequences on public acceptance, ageing management shall primarily focus onthe prevention, which is the first step of the defence-in-depth approach.

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The achievements of the 3-rd and 4-th Euratom Framework Programmes have beenpresented in [92] and [95], together with the orientations of the above 5-th programme:Topics covered are illustrated by Figure 2.8.

As far as plant safety is concerned, the 3-rd Framework Programme was focussedparticularly on severe accident evaluation, these actions being continued under thecontext of the 4-th programme together with actions on ageing and probabilistic safetyanalyses. This differs from the 5-th programme, which is mainly focussed to problemssolving and socio-economic challenges.

What was new with the 4-th programme was its consistency with the 1992 MaastrichtTreaty stating the obligation to co-ordinate research and technological developmentactivities so as to ensure consistency between national policies and community policy.This has led to an increased weight of consultative committees with representativescoming from governments, regulatory authorities, manufacturing industry, electricutilities, universities, research organisations and others.

Reference [92] is mainly devoted to actions concerning severe accidents, whichessentially concern new projects with evolutionary design. Concerning ageing actions,the 4-th Framework Programme includes 11 actions mainly on structural (essentiallymetallic) components subjected to irradiation embrittlement or other operatingconditions.

Three projects addressed neutron irradiation and dosimetry:

− AMES has established the dosimetry to be used as well as the irradiationprogrammes to be carried out under this network,

− MADAM has reviewed the neutron damage indexation and prepared conversiontables of damage indices used for embrittlement assessment,

− RESQUE addressed the issue of qualification, verification and comparison of studand electron beam welding for reconstitution of irradiated impact and toughnessspecimens.

Concerning neutron irradiation induced material degradation, the REFEREE projectdeals with the experimental determination of the difference between toughness andCharpy impact transition curves.

With respect to structural integrity of components and welds, three projects are co-funded by the European Union:

− DISWELD deals with techniques for assessing environmentally assisted crackingof dissimilar metal welds,

− BIMET carries out pipe tests to follow crack growth and path of cracks by variousanalysis methods,

− VORSAC studies the evolution of residual stresses and related phenomena

39

)LJXUH�� Schematic representation of all FP-4 and FP-5 projects in thefield of plant life extension and management, using the structureof FP-5 (extracted from ref. [95]).

Four actions concern more strategic studies reviewing the range of ageing relatedtopics relevant to nuclear safety, which constitute the essential task of the presentstudy. Sate of the art reports are being prepared on the following topics:

− MODAGE action concentrates on the present capabilities of modelling corrosion inprimary and secondary circuits,

− INTACT surveys ageing research activities and will indicate further researchneeds,

− AMES-NDT reviews NDT and NDE techniques that are or will soon be available,

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− EURIS analyses the tools available for defining in an effective way risk-informedin-service inspection programmes [91].

Containment integrity actions have been identified, which mainly deal with behaviourunder accident conditions.

A report has been issued on the evaluation of nuclear expertise in Europe [93]. Theintent of this report is to collect and analyse information on nuclear expertise and tomake recommendations enabling to maintain a high safety level in the existinginstallations and to keep the nuclear option open. 180 answers were received fromvarious experts.

Ref. [92] summarises the key items of this study, in particular the need for a sufficientlevel of expertise to run safely the current generation of reactors and to ensure thenecessary safety margins in ageing reactors. Especially in the case of plant lifeextension, one has to understand the meaning and intent of existing safety margins,before "eating" them. This leads to keep a minimum of nuclear educated engineers ableto think and to act beyond standard operating procedures.

The report also notices that due to lack of student interest, teaching staff and universitylaboratories are ageing (and the same may become true for safety authority staff itself).

It was even said that "a bad reactor with a good team is usually better than a goodreactor with a bad team". This point has led to some of the recommendations includedin the 5-th Framework Programme.

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One of the overall objectives of the Community’s activities is to help harmonise thenational safety rules and regulations. In order to achieve its strategic goals in the areaof nuclear energy, DG ENV is running the "Nuclear Regulator’s Working Group"(NRWG), made of representatives from safety regulatory authorities from EU MemberStates and Applicant States.

The same organisations plus nuclear operators from the same countries, who weremembers of the "defunct" Reactor Safety Working Group discontinued in November1998, are now members of the new European Nuclear Installations Safety Group(ENIS-G). This group will have to deal with the general framework to providemeasures dealing with sharing of experience, particularly in issues such as ageing orobsolescence of some components.

The NRWG provides a forum for exchanges covering in particular the following topicswhich are of interest for ageing management:

− feedback of operating experience,− risk analysis with emphasis on risk based in-service inspection methods,− operational safety fundamentals and development of safety culture.

At the Industry level, expert groups have been established, which inscribe theiractivities within the general orientations defined in the above-described programmes.As far as ageing is concerned, reference [59] is particularly dedicated to a preparatorywork for an indicative programme related to ageing issues to be handled by theWorking Group on Codes and Standards, presented in Section 2.2.3 below.

It includes a synthesis of existing work made at the international level, with 142references identified. The status of the ageing research programmes at theinternational level and in the various countries is addressed.

41

The R&D effort on ageing is analysed in this report, including work recently carriedout, on-going programmes, and future activities. Gaps in the knowledge which need tobe covered in future research have been identified, including:

− Life assessment technology and NDE for physical properties altered by ageing− Improved monitoring techniques,− Reproduction of phenomena on test specimens,− Research on reactor pressure vessel internals degradation and mitigation,− Repair and degradation mitigation technologies.

In order to perform this work, the report proposes a tentative of programme, groupedinto six topics:

− Database generation (toughness, mechanical properties of aged materials…),− Material degradation studies (typical failure selection, detailed examinations,

verification of code and standard validity),− Degradation detection techniques (toughness testing, direct evaluation of actual

material properties, small specimen testing, correlation between KV tests andtoughness properties, development of new monitoring systems),

− Assessment procedures (conservatism and improvement of prediction formulas,determination of margins which guarantee safe operation, evaluation of nocivity offactory defects),

− Repair/mitigation technologies (improved materials, stress relief, repair procedures– under-water, robotics – optimised water chemistry)

− Recommendations for Codes and Standards (introduction of actual data instead ofdesign data, material characteristics curves, toughness determination from KVspecimens, standardised and validated NDE methods to determine physical andmechanical properties, recommendations on surface conditions).

Among the various international studies, report [59] also summarises the contextswhere ageing studies are covered in the European Union:

− The Working Group on Codes and Standards (WGCS),− The Ageing Materials Evaluation and Studies (AMES),− The European Network for Inspection Qualification (ENIQ),− The Network for Evaluating Steel Components (NESC).

The group activities are described hereafter.

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The objective of the working group on codes and standards, initially created in thecontext of the European Fast Reactor development, and enlarged later to cover LWRactivities is to analyse the existing codes and standards in order to identifydiscrepancies and to carry out studies facilitating the reduction of this discrepancy bythe member states, allowing a pre-harmonisation of European codes and standards.

The WGCS is not specifically dedicated to ageing evaluation. It also covers design andconstruction rules applicable at the plant design stage.

An overview on pre-harmonisation studies conducted by WGCS has been published in[83].

The WGCS organisation includes three activity groups: AG1, AG2, AG3, the activitiesof which are presented below, as far as they concern ageing management:

42

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AG1 activities cover in-service surveillance, including non-destructive examinationand material degradation evaluation activities in the general framework of in-serviceinspection (ISI) activities. Defect assessment and evaluation of flaws also fall in itsscope as well as repair, maintenance and component replacement practices.

In the context of these activities, the non-destructive examination (NDE) practices andassociated acceptance criteria given in the various ISI codes (essentially RSE-M [60]and ASME XI [61]) have been evaluated [62]. It was found that these codes, thoughdifferent with regard to their detailed rules and allowable defect tables, providecomparable intrinsic safety coefficients for the various operating conditions.Differences found may be related to differences on the consideration of in-servicecrack growth potential.

Progresses in the harmonisation of European in-service inspection codes have beencovered in report [63]. Two parts deal respectively with LWR and FBR. ConcerningLWR, the major differences in the requirements of RSE-M, KTA [64] and ISIspecification for Sizewell B based on ASME XI with attachment specifications agreedwith Nuclear Installation Inspectorate (NII) are identified. The report concentrates ondifferences in the ISI of key primary circuit components (RPV, SG, pressuriser, reactorcoolant pumps and pipework).

The intent of NDE is to find potentially detrimental defects. Qualification of NDEtechniques consequently depend on the suspected defects. According to RSE-Mclassification:

− conventional qualifications are applied where no defect is suspected (this applies atthe construction stage, and also applies to the majority of in-service examinations)

− general qualifications are applied where the presence of particular defects ispostulated, following return of experience on similar plants,

− specific qualification, which shall be obtained on realistic defects is required wherethe presence of particular defects is stated for the construction under consideration.

The report [65] deals with the comparison of the various practices in this field andcovers the construction and operation stages. It also refers to Nuclear RegulatorsWorking Group (NRWG) document on qualification of NDT [66], and to ENIQ ReportN°2 [67].

Other reports, such as [68], extend the comparison of ISI practices to eastern countries,and reports [69] and [70] cover the relationship between manufacturing standards andin-service inspection needs, i.e. inspectability.

Continuous on-line monitoring of NPPs components has also be covered in report [71],which cover, in particular:

− the identification and comparison of Codes and Standards including continuousmonitoring requirements. Areas where harmonisation may be desirable aresuggested,

− the survey of continuous monitoring techniques for structural damage and materialdegradation detection. A list of the major degradation mechanisms for various NPPcomponents has been compiled together with a list of techniques which may beused to monitor each mechanism.

The importance given to ISI code, defect acceptability and on-line monitoring is linkedto the fact that no real confidence can be given to ageing management processes if theycannot be based on real and efficient examinations conducted on actual components.

43


AG2 activities cover design and analysis topics. Its role was enlarged to cover thestrategy to be followed for the future evolutions of the design and construction codes,taking into account the new European projects and the transposition of the EuropeanPressure Equipment Directive [72], which is not strictly speaking applicable to themost important nuclear components, but which will influence the overall standardslandscape.

AG2 activities are not specifically dedicated to ageing evaluation, but the toolsdeveloped for a better harmonisation of practices, particularly in the fatigue and fastfracture evaluation domain may be used in the context of the study. In particular, thefollowing topics have been covered in AG2 reports:

− fatigue and creep-fatigue damage assessment, including syntheses on fatigue dataand parameters likely to influence fatigue resistance [73], or benchmarks for theevaluation of thermal striping in order to explain actual phenomena and to permitmore realistic damage prediction [74].

− fast fracture evaluations, which needs a good understanding of the conditions oftransferability of data obtained from specimens to the analysis of defect assessmentin real structures [75], [76], [77], and the evaluation of the effect of potentialresidual stresses [78]. In addition, acceptance of defects have to consider theirpotential behaviour under faulted conditions: report [79] evaluates in particular thedefect tolerance under conditions for which ASME or RCC level D [94] criteria areapplicable.


AG3 activities deal with code harmonisation and development of expertise in the fieldsof materials selection and properties. This includes in particular:

− evaluation of in-service degradation of material properties by the effect of neutronirradiation, corrosion or thermal ageing,

− evaluation of properties of materials, with special emphasis on fracture mechanicstesting procedures and fatigue properties.

Work on re-evaluation of toughness reference curves for fracture mechanics analyseshas in particular been done [80], and environment effects on fatigue are currentlyreviewed. Nevertheless, there are few studies specifically dedicated to ageing, which iscovered by AMES Reports (see section 2.2.4 below).

A report [81] has made the synthesis of the results of the RPV irradiation surveillanceprogrammes of four countries and has compared these results with the predictedcalculated values.

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Adequate understanding of the mechanisms that can contribute to component failurecan only be gained through supporting research programs. R&D institutions in Europehave consequently joined their efforts in European Networks to deal with the followingproblems, which have been considered important:

− understanding and modelling of the major ageing phenomena and validation ofmitigation measures,

44

− research and development of techniques for on-line monitoring of thesedegradations,

− improvement, better use and qualification of periodic inspection procedures,− capability of precise structural assessment to evaluate safely the remaining margin

to failure.

Each of these networks: ENIQ, AMES, NESC deals with a specific aspect of thefitness for purpose of materials in structural components.

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The European Network on Ageing Materials Evaluation and Studies (AMES) was setup to bring together the organisations in Europe that have the main capabilities on RPVmaterials assessment and research, with the following objectives:

− Establish and execute AMES projects in this area,− Act as an European review Group,− Provide technical advice to regulatory bodies, General Directorates of the EU and

provide a base for the development of common European standards,− Participation in collaborative programmes with the Russian Federation and Eastern

Europe.

Activities include:

− review of capabilities within member organisations and existing knowledge base,− study of RPV and other components, such as reactor internals,− assessment of the availability of stocks of irradiated and un-irradiated materials

including those that might be recovered from operating or decommissionedreactors,

− understanding of the underlying effects for irradiation damage, thermal ageing andannealing,

− annealing validation an re-irradiation studies of LWR materials,− the survey of national regulatory requirements and standards at European level

relevant to material damage and mitigation methods.

Various reports have been issued, particularly on irradiation embrittlement mitigation[87] and thermal annealing [88]. AMES has developed a strategy for subjects related toageing [86].

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The objective of ENIQ network is the co-ordination and management at Europeanlevel of expertise and resources for the assessment and qualification of NDE inspectiontechniques and procedures, primarily for nuclear components. The primary focus is onISI performance demonstration.

ENIQ proposals is the basis for establishing a European attitude about inspectionqualification in general.

Three major tasks have been covered:

− Gathering of information which can be of interest for inspection qualification:correlation between real and realistic flaws, inventory of available assemblies andblocks, application of human factors and reliability studies, structural integritysignificance of flaws.

− Studies on qualification schemes of ISI procedures, development of qualificationprocedures for specific reactor components, development of a simulator forultrasonic inspections. "Recommended Practices" have been defined, which are

45

considered for defining NDE qualification requirements in codes and standards,such as RSE-M.

− Application of what is acquired within the first groups of tasks: managementschemes of available resources, development of accreditation criteria, co-operationwith eastern countries.

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The assessment of integrity of a structure containing flaws is an interdisciplinarityactivity with inputs from NDE, materials testing, stress and fracture analysis. Severalcollaborative programmes have been conducted in these various topics as shown aboveand also in PISC (Programme for the Inspection of Steel Components) [84] andFALSIRE [85] concerning respectively NDE reliability and fracture mechanics areas.

It is the interaction between the disciplines which is essential in order to take righttechnical decisions and judgements. The concept of NESC is consequently to exploreinternational practices in the entire process of structural integrity assessment. Particularattention has been given to:

− capability of current NDE techniques and interpretation of NDE data− statistical treatment of NDE data for structural assessment− allowance for stable tearing and warm-prestressing− analysis of complex stress fields for fracture assessment,− integrity arguments and risk evaluation,− design of large scale structural test and validation of assessment methods on full

size components.

The experimental programme investigates the pressurised thermal shock transient bymeans of the procurement and fabrication of a cylinder with cracks and materialsproperties to simulate a damaged pressure vessel, the inspection of this cylinder byparticipating organisations using various ISI tools, integrity assessment and test of thesimulated vessel in the spinning cylinder test facility at Risley. The test is followed bypost-inspection using ISI procedures, reassessment of structural integrity using newNDE information, destructive examination, and re-evaluation of NDE and structuralintegrity assessments.

This approach allows to evaluate the global consistency and pessimism of the variousmethods and hypotheses.

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As shown in this report, various aspects of ageing management are addressed indifferent European programmes and contexts.

WGCS, AMES, ENIQ and NESC studies provide valuable inputs for the evaluation ofthe ageing of mechanical components, as well as in certain cases a complete strategyfor the consideration of particular ageing phenomena, such as RPV embrittlement.

Topics retained in these studies resulted from a consensus among the variousparticipants, which did consider prioritisations coming from the return of experience orfrom studies made in other contexts. Except in ref. [59], no strategy documents wereprepared in the field of ageing management, which is the reason why the present studyhas been decided.

46

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�� ,1752'8&7,21Many studies are under way in the energy production industry to maintain high plantcapability and availability factors in face of ageing phenomena. Study of suchphenomena and development of mitigation measures is thus part of the operation ofpower plants.

Different ageing phenomena can contribute to degradations that can lead to extendedplant outage or shorten the projected lifetime of the plant. Typical failure mechanismsinclude:

− corrosion and stress-corrosion cracking,− irradiation and thermal embrittlement,− vibration, thermal cycling, fatigue,− wear-out, erosion, deposit effects,− creep in fast breeder reactors and some components of non nuclear plants.

These phenomena have to be studied and understood, and in several cases, adequateunderstanding of the mechanisms can only be gained through supporting laboratorytests research programmes.

A plant life management programme shall include the following tasks:

− identification of components of importance to plant life,− identification of life threatening factors,− understanding and description of ageing phenomena,− development of methods for in-service ageing detection,− identification of mitigation methods,− evaluation of components replacement possibilities,− study of annealing and repair methods for non replaceable components.

This Section describes the main ageing phenomena which may occur in a NuclearPower Plant (NPP). The methods of identifying ageing in service and mitigation ofageing are also covered.

Ageing phenomena can be physical (resulting in deterioration of physicalcharacteristics of Systems, Structures and Components (SSC)). If ageing is noteffectively managed, it could affect the capability of SSC to perform their requiredfunctions.

They may also be non-physical. Paragraph 3.3 covers these other aspects, whichinclude human ageing and ageing of culture and procedures.

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The list of main degradation mechanisms likely to affect structures and components isgiven in Table 3-1 below. Potential ageing mechanisms identified for the various typesof components are listed in Appendix 3.

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Materials exposed to neutron radiation undergo changes in microstructure andproperties. The extent of the changes depends on the material, the neutron flux, the fluxspectrum, the fluence, and the irradiation temperature.

47

The reactor pressure vessel, made in Low Alloy Steel, is subjected to neutronirradiation in the core region, which results in progressive embrittlement. Thisembrittlement results macrostructurally as hardness and yield stress increase, and afracture toughness decrease.

Fracture toughness is the property that governs the material resistance to fast fracture,which, for ferritic steels, is small at low temperature, where the material behaves in abrittle manner, and increases with the temperature until the material becomes ductile. Itis generally accepted that the effect of irradiation is to shift this curve to highertemperatures, the shape of the curve being only slightly affected, with a small decreaseof the upper shelf toughness in the ductile regime.

Also organic materials (e.g. electrical cables) degrade due to irradiation embrittlement.

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1 Irradiation X X X2 Thermal ageing X X X3 Creep X4 Fatigue X

4,1 High Cycle Fatigue X4,2 Low Cycle Fatigue X4,3 Thermal Fatigue X

5 Corrosion X X X5,1 Corrosion without mechanical loading X

- Uniform Corrosion Attack X X- Local Corrosion Attack (Pitting, Wastage, Crevice Corrosion) X X X- Selective Corrosion Attack (InterGranular Corrosion) X

5,2 Corrosion with additional mechanical loading X- Stress Corrosion Cracking (InterGranular SCC, TransGranular SCC) X X- Primary Water Stress Corrosion Cracking X- Hydrogen Induced Stress Corrosion Cracking X- Strain Induced Corrosion Cracking X- Corrosion Fatigue X

5,3 Flow Accelerated Corrosion (Erosion/Corrosion) X5,4 Irradiation Assisted Stress Corrosion Cracking X5,5 Microbiologically Influenced Corrosion X

6 Wear (Fretting, Abrasion, Vibration, Cavitation, ...) X X7 Loss of prestressing X8 X

- Freeze-Thaw Cycling, Wetting and Drying X- Chemical Attack X- Oxidation X

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Potential degradation for

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Environment effects

Concrete Degradation (Shrinkage, Leaching of Calcium Hydroxide, Reaction with Aggregates)

48

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Thermal ageing refers to gradual and progressive changes in the microstructure andproperties of a material exposed at an elevated temperature for an extended period oftime.

For Ferritic Steels, the effect of thermal ageing results mainly from intergranularsegregation of residual impurity elements (e.g. Phosphorous). This leads to a hardeningof the material and a reduction of the fracture toughness.

Cast Austenitic-Ferritic Stainless Steels (duplex structures) and Martensitic StainlessSteels are susceptible to thermal ageing in the normal operating temperature range ofPWRs., by "unmixing" of chromium from its solid solution in the ferrite of the duplexstructure and in martensite [102].

Also organic materials (e.g. electrical cables) degrade due thermal ageing leading tochanges in mechanical properties (e.g. hardening) and cracks initiation.

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Creep is the plastic deformation that occurs over a period of time in a materialsubjected to a stress, even below the elastic limit. It is considered to result from thecompeting process of work hardening and thermal recovery. The process is thenthermally activated and is significant above 0.4 times the melting point of the material.

Creep is a potentially significant ageing mechanism for pre-stressed or post-tensionedstructures.

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Fatigue is produced by periodic application of stresses by mechanical or thermalloading. The metal subjected to fluctuating stress will fail at stresses much lower thanthose required to cause fracture in a single application of load. The key parameters arethe range of stress variation and the number of its occurrences. Technologicalconditions (i.e. surface roughness and residual stresses) and Environmental conditions(presence of deleterious chemical species) may also play a role.

Low cycle fatigue, usually induced by mechanical and thermal loads, among whichgamma-heating, is distinguished from high cycle fatigue, mainly associated withvibration or high number of small thermal fluctuations.

At the design stage, all Class 1 components were qualified for stress and fatigue usingthe so-called "design transients" according to the design code. A re-assessment of someof these components may be necessary for one of the following reasons:

− the numbers of design transient are associated with a predetermined design lifetimeof the components (30 or 40 years);

− in the older plants, the stress and fatigue analyses were performed with outdatedtechniques;

− differences between operation transient book keeping and envelope designtransient,

− the design transient may be modified due to component replacement (e.g. SG),NPP uprating, or new operation modes (e.g. extended fuel cycle);

− newer version of the code may have different requirements (e.g. on materialproperties, on fatigue analysis),

− unexpected transient may show up (e.g. thermal stratification transients in the surgeline or in the feed water lines, thermal transient in unisolated sections of pipingconnected to the reactor coolant system); and

49

− at the time of design, some components may be covered by a different design code(e.g. certain Class 1 pipes analysed according to the ANSI B31.1 Code which doesnot require explicit fatigue qualification).

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Corrosion without mechanical loading

Corrosion is the degradation of a material by chemical or electrochemical reaction withits environment. There are many forms or effects of corrosion, depending on thematerial and environment. Corrosion is characterised by material loss or deteriorationof its mechanical properties. Corrosion reduces the component wall thickness, eitheruniformly or locally.

General corrosion refers to a uniform attack over surfaces of the material and results inthinning of the material. (e. g. Corrosion of Carbon or Low Alloy Steels exposed toboric acid leakage). General corrosion rates vary with fluid oxygen content,temperature, flow rate.

Localised corrosion includes pitting, crevice corrosion, etc. Pitting is commonly causedby the breakdown of the passive film on a metal, in local areas, by species such aschlorides. Crevice corrosion results from local environment conditions in the restrictedregion of a crevice being different and more aggressive than the bulk environment.Pitting may occur at critical location such as tube-to-tubesheet joints of the SteamGenerators, SG, where fluid velocities are stagnant or low (i.e. impurity concentrationcan occur).

Stress Corrosion Cracking

Stress Corrosion Cracking, SCC, is a localised non-ductile failure which occurs onlyunder the combination of three factors: tensile stress, aggressive environment, andsusceptible material. The Stress Corrosion Cracking failure mode can be intergranular,IGSCC, or Transgranular, TGSCC. In a NPP, Primary Water Stress CorrosionCracking, PWSCC, and Irradiation Assisted Stress Corrosion Cracking, IASCC, arealso defined.

Intergranular Stress Corrosion Cracking is associated with a sensitised material (e.g.Sensitised Austenitic Stainless Steels are susceptible to IGSCC in an oxidisingenvironment). Sensitisation of unstabilised austenitic Stainless Steels is characterisedby a precipitation of a network of chromium carbides with depletion of chromium atthe grain boundaries, making these boundaries vulnerable to corrosive attack.

Transgranular Stress Corrosion Cracking is caused by aggressive chemical speciesespecially if coupled with oxygen and combined with high stresses.

Primary Water Stress Corrosion Cracking is a form of IGSCC and is defined asintergranular cracking in primary water within specification limits (i.e. no need foradditional aggressive species) (e. g. IGSCC of Inconel 600 Alloy in primary water).

Irradiation Assisted Stress Corrosion Cracking

Irradiation Assisted Stress Corrosion Cracking refers to intergranular cracking ofmaterials exposed to ionising radiation. As with SCC, IASCC requires stress,aggressive environment and a susceptible material. However in the case of IASCC, anormally non-susceptible material is rendered susceptible by exposure to neutronirradiation. IASCC is a plausible ageing mechanism for PWR internal components(e.g. baffle bolts).

50

Flow Accelerated Corrosion

Flow Accelerated Corrosion, FAC, also called Corrosion-Erosion mechanisms, refersto the combined action of erosion (i.e. the mechanical action of a fluid on a metalsurface) and corrosion. The severity of erosion vary with the material type, the fluidtemperature, the fluid velocity, the oxygen content in the fluid, and the componentgeometry. The result of FAC is an increased rate of attack on metal because of therelative movement between a corrosive environment and the metal surface. Carbon andLow-Alloy Steels are susceptible to FAC.

Another erosion process usually experienced on pump impellers is cavitation.Cavitation is a process that occurs when the local pressure in a flowing liquid isreduced without a change in temperature. Hence, vapour-bubbles form. within theflowing liquid. When these bubbles pass into a region of higher pressure, they collapse,producing high-localised fluid velocities and causing damage to material surface.

Microbiologically Influenced Corrosion

Microbiologically Influenced Corrosion, MIC, is the accelerated corrosion of materialsresulting from surface microbiological activity. MIC is characterised by the formationof microbial colonies and associated scale and debris on the surface of the metal. MICaffects Carbon Steels and, to a lesser extent, Stainless Steels and Nickel alloys.

Any buried system or system using untreated water is susceptible to MIC. The majorfactors influencing the growth of MIC are temperature, pressure, pH, water content,and oxygen.

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Wear is defined as a loss of material as a result of mechanical contact between twosolid surfaces due to vibration (e.g. wear was experienced on SG tubes due to contactwith anti-vibration bars), sliding, or due to the presence of loose/foreign objects. It isreferred as Fretting Wear if the surfaces are in the presence of a corrosive environment.

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A material loaded to an initial stress may experience a reduction in stress over a periodof time. Stress Relaxation occurs under conditions of constant strain. Stress relaxationis accelerated at elevated temperature and/or in the presence of fast neutron irradiation.

Typical components subject to a loss of prestressing are the prestressed cables in theprimary containment.

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The ageing mechanisms of the structure – such as fatigue resistance – depend onenvironment. Various ageing mechanisms may also act simultaneously, like fatigueand creep, for example.

The environment by itself may also cause the degradation of the structure, such aswetting-drying and freeze-thaw cycling, chemical attack, etc.

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Concrete SSC may be affected by several ageing mechanisms resulting in an alterationof the mechanical properties or the physical form of the concrete WKXV�DIIHFWLQJ�WKHstructural integrity of the SSC (e.g. aggressive environment can increased porosity,perrneability, and reduced concrete strength; atmospheric carbon dioxide can causeconcrete carbonation reducing the properties of the concrete; flowing water overconcrete surfaces can remove significant amount of concrete).

51

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All structures settle to some extent during and after construction depending on thephysical properties of the foundation material. Severe settlement can cause concretecracking and/or misalignment of equipment and can lead to high stresses within thestructure.

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Various approaches may be used to identify ageing in service, depending on ageingphenomena:

• Some are anticipated and lead to a specific in-service surveillance program: this isin particular the case of the neutron embrittlement of the reactor pressure vesselcore shell,

• Others are not anticipated or are subjected to precautionary measures in order toavoid them: this is the case, for example, of the fatigue damage, for which detailedcalculations are done and in-service transient follow applied in order to ensure thatthey remain within the design hypotheses. In this case nevertheless, periodic in-service examinations are conducted to verify if non-anticipated damage maynevertheless appear. These examinations may depend on the margins demonstratedduring the above evaluations,

• Notwithstanding the above precautions, damage risks may appear during service,leading to leaks or failures. They may result from non anticipated phenomena (forexample thermal ageing of cast parts), or from under-estimated phenomena (forexample, fatigue under vibration or thermal transients). This leads to a progressiveimprovement of the construction choices and a complementary specific inspectionprogram for existing plants likely to be affected by the phenomena.

The physical methods used for ageing evaluation may concern:

• Material mechanical and/or physical properties measurements,• Component integrity,• Component functionality.

These various concerns are developed in the following paragraphs.

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The best known material surveillance program is the one dedicated to the neutronembrittlement of the reactor belt line. Specific samples are placed between the core andthe vessel and are tested according to a planning defined according to embrittlementprevision rules, permitting to anticipate the evolution likely to affect the vessel shellitself.

Internal procedures or design codes give acceptable methods for establishing theseprograms, such as US Regulatory Guide 1.99 [103], [104], KTA [64], or ETC-M [105]standards.

These programs include material test samples (tension and toughness specimens),neutron flux monitoring dosimeters, and thermal monitors.

Such programs include provisions on material selection and test specimen sampling,capsule filling and positioning, capsule removal schedules, and eventual referencematerials for comparisons or identification of particular operating conditions.

Where other significant material ageing phenomena are anticipated, such as thermalageing, a dedicated program may be required, in particular by the new French

52

"operation" Order dated November 10, 1999 [106]. Such programs do not necessarilyimply the introduction of material samples within the component, but may includeartificial ageing tests conducted under representative conditions.

When a specific ageing effect is discovered, test samples may be taken from actualcomponents in order to confirm the hypotheses of the generic justification files.

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The industrial and regulatory practices include periodic integrity tests, generally at aeight or ten years interval. At the occasion of these tests, or between two tests, aninspection program is applied, involving visual and volumic non-destructiveexaminations, which are described in codes, such as US ASME XI [61], German KTA[64], or French RSE-M [60].

In addition to these periodic "re-qualification", additional reduced examinations areconducted at each refueling shutdown.

These code requirements are completed by particular examinations where specificageing phenomena are anticipated, in particular following return of experience. Theseexaminations are depending on the importance of the phenomena for the safety orintegrity of the component, and they may concern primarily "precursor" components(even when they are not important from a safety point of view) in order to anticipatethe phenomena likely to affect the most important components.

The content of the inspection programs are modulated by the results of design studies,such as fatigue or fast fracture evaluations, some zones appearing more "sensible" to agiven degradation phenomena than others. Nevertheless, the conclusions of theseevaluations are completed by a minimum number of additional "random" examinationsin order to detect non anticipated phenomena. The inspection strategy also depend ifthe plant under consideration is part of a series or not.

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Component performance is periodically evaluated through functional tests. Thisconcerns primarily active components such as pumps and valves, or snubbers andelectrical equipment. Loss of prestressing or containment leaktightness are alsosubjected to periodic measurements.

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Various destructive tests and expertises and non-destructive examination methods maybe used for the evaluation of the physical ageing of materials and structures. They aresummarised in Table 3.2.

Some of the methods are still under development. This is in particular the case forDNPA, PTE, Barkhausen or annihilation of positrons effects. DNPA method wasnevertheless used in the generic evaluation of thermal ageing of Austenitic-Ferritic castelbows in France [107].

In addition, some of the methods do not give an absolute evaluation of ageing, but arelative information, implying that an initial measurement has to be made, whichmeans that the ageing effect has to be anticipated.

Detection of damage initiation may be obtained by the use of NDE methods allowingthe detection of the corresponding cracks or by an indirect detection of theconsequences of the existence of such defects, for example the acoustic emission of thematerial in the vicinity of these flaws, when subjected to loading. The interest of suchan approach is to identify and locate damage, even in zones where it was notanticipated.

53

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Irradiation embrittlement - Mechanical test samples taken according to given schedule- Hardness tests- Annihilation of positrons

Thermal ageing - Mechanical test samples taken from actual components andexpertise of replaced components

- Hardness and micro-hardness tests- Small angles neutron diffraction (DNPA)- Thermo-electric power (PTE) based on Seebeck effect- Simulation in representative laboratory conditions

Fatigue initiation phase - Measurement of residual stresses and strain-hardening by X-ray diffraction

- Eddy current- Barkhausen effect- Infra-red thermography- Fatigue samples witnesses

Detection of fatigue cracks orCorrosion (intergranular, stress-corrosion) cracks

Non-destructive examinations:- Surface examinations: visual or remote (TV)

liquid penetrantmagnetic particle

- Volumetric examinations: Eddy currentradiography and gammagraphyultrasonic

Acoustic emissionLeak detection

Uniform and local corrosion,Wear

Thickness measurements by ultrasonic examination

Loss of prestressing Strain and deformation measurements by means of strains gaugesand dynamometers

Concrete degradationDifferential settlement

- Visual inspection- Topographical levelling- Fissurometers- Leaktightness tests

Electrical cables - Monitoring of mechanical and electrical characteristics- Qualification under representative conditions- Visual inspection

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The effects of ageing on SSCs (Structure, System and Component) is detected throughdifferent mechanisms in Plant operation (preventive and corrective maintenance,inspections, monitoring, SSC performance monitoring). The mechanism that has givenrise to the degradation of the behaviour or characteristics of SSC material can bedetermined through the subsequent analysis of these effects. The election of anefficient method for mitigating ageing in SSC material depends on an accuratedetermination and evaluation of the degradation mechanism that caused the ageing.

54

Table 3-3 contains the most usual ageing mitigation methods for SSCs in nuclearpower plants. These mitigation methods can be grouped into three main categories:

− Change of SSC Design− Change/Recovery of Material Characteristics− Changes of Operating Parameters

The following subsections describe these groups and their application to the differenttypes of SSCs in a NPP.

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This group of mitigation method includes design modifications on components,systems and structures which do not involve a complete change or replacement butmakes them immune to a degradation mechanism or minimises the effects of ageingcaused by said mechanisms.

In piping systems where Erosion/Corrosion has been detected, the definition of a newlayout of specific piping sections minimises or even prevents this degradationmechanism. The purpose of this redesign is to obtain layouts where areas with highturbulence are avoided. In wet steam systems (two phases) it is usual to redesign thelocation of traps in order to ensure moisture contents of less than 2%. These designchanges in piping layouts normally go together with material changes, replacingcarbon steel sections with alloyed steel of at least 2.5% Cr, or stainless steel.

In steam generators fretting wear appears in the tube walls in the curved area caused bytheir coming into contact with the Anti-vibration Bars (AVB). The mitigation of thisdegradation mechanism involves the modification of the design of the anti-vibrationbars. This design modification is normally aimed at: 1) Reducing the gap between thetubes and the AVB in order to prevent badly supported tubes; 2) Increasing thetube/AVB contact area; 3) Changing the AVB material in order to improve thecoefficient of wear of the tube material.

Old plants with high cycle fatigue induced by vibration have modified their downwardbypass flow scheme into an upward bypass flow scheme. In new reactors, the bypassflow enters the core baffle-core barrel region from the bottom region of the core barrel,and the bypass flow travels upward through holes in the former horizontal plates.

Electrical components (motors, cable, MCC, chargers, inverters, etc) are subjected tothe effects of thermal ageing during their operating life. Reanalysing their operatingand ambient conditions based on historical data may give rise to design modificationsto optimise the cooling and ventilation systems of the components and the room inwhich they are installed.

55

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The most common evidence of ageing in a component or structure is the degradation ofthe physical characteristics of its material. Consequently, a large number of mitigationmethods are aimed at the recovery or improvement of material characteristics.

Intergranular Stress Corrosion Cracking (IGSCC) is a degradation mechanism thatfrequently occurs in the recirculation lines of BWR plants. Different mitigationmethods, defined according to the stressor to be eliminated, have been used for thisdegradation mechanism. When the objective is to minimise the effect of fluid on thematerial, the Corrosion Resistant Cladding (CRC) method implies applying a metallicprotective coat on the inside of the piping to protect the material’s sensitised area. Inother IGSCC mitigation methods the stressor to be minimised or eliminated is materialsusceptibility. An example of this method are solution treatments by means of whichthe sensitised area is heated and then subjected to quick cooling to retain carbon. Whenthe purpose of the IGSCC mitigation is to eliminate tensile stress, Weld Overlay StressImprovement (WOSI) and Induction Heating Stress Improvement (IHSI) throughexternal heating while passing cold water through the piping provides compressionresidual stresses.

Reactor vessel annealing is the method used to recover the loss of vessel materialtoughness due to the effect of neutron embrittlement. Standard ASTM E 509-86 “In-service Annealing of Light Water Cooled Nuclear Power Reactor Vessel” [109]establishes the general procedures for carrying out LWR vessel annealing and ensuringits effectiveness.

This group of mitigation methods also covers all those that involve the application orrenewal of the material’s protective coating or lining with the aim of mitigatingerosion, corrosion and erosion/corrosion. This includes: application of protective paintinside the service water and circulating water system piping that usually carry rawwater; application of stainless steel weld overlay on piping elements affected byerosion/corrosion; installation of sleeves in the tubes of the steam generator and otherheat exchangers; application of metallic coating on tanks that are vulnerable tocorrosion.

The mitigation of concrete structures is aimed at restoring the integrity of the affectedcomponent and stopping the degradation mechanism, ensuring as much as possible thatthe cause of the degradation is not repeated. The mitigation methods are definedaccording to the degradation effect observed (cracking, spalling, porosity). Theobjective of methods aimed at repairing and mitigating cracking is to prevent the entryof elements that are harmful to the structure. Depending on the characteristics of thestructure and damage caused, these methods include: epoxy resin injection, crackstapling, flexible sealing, slurry injection and polymer soaking. Spalling and scalingmitigation methods consist in cleaning the surface, applying fixing agent, applyingcement, epoxy resin or high quality concrete. Porosity is repaired by drilling small-diameter holes that intercept the pore, cleaning with compressed air or water, and theninjecting epoxy resin, slurry or epoxy foam, depending on the size of the pore.

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For some components and specific degradation mechanisms affecting them, it ispossible to optimise component operation, by means of operating parameters control,in order to get a mitigation of the degradation phenomena incidence. This strategy ispossible only if the degradation has not risen to certain level for which the componentmust be repaired or replaced.

Pressure, temperature and fluid chemistry control, among limit values previously fixed,are the most used method in order to mitigate the degradation during operation. In

56

general, parameters to be controlled and limit values are depending on the dominantdegradation mechanism for each component, maintaining a balance betweendegradation minimisation and maximum performance.

The main degradation mechanisms which ageing effects could be reduced by means ofoperating conditions optimisations are basically fatigue and corrosion.

Following the three main methods regarding operating conditions optimisation areincluded and explained in more detail.

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This mitigation method is basically used to reduced the effects of corrosion. It is wellknow that corrosion, and specifically stress corrosion cracking, is produced as aconsequence of three causes: susceptible material, stress and aggressive environment.Therefore, the corrosion effects and incidence could be reduced avoiding or controllingthe above mentioned causes, in particular reducing the environmental aggressivity andlimiting the accumulation of corrosive products, pH, etc.

Following, there are some of most common methods used in reactor pressure vessel,steam generator, piping, internals and heat exchangers.

Control of secondary water chemistry

Use of AVT (All Volatile Treatment, PWR). This method is applied to reduce, amongother effects, the incidence of wastage/thinning in steam generator.

Control of primary water chemistry

Use of H2 and LiOH for oxygen and boric acid control. In pipes, the hydrogen injectionis used to control oxygen concentration in the range of 10 to 20 ppb.

Addition of products to maintain the materials on suspension (dispersion) and avoiddeposition.

Addition of inhibitors (phosphates, nitrites, etc.) and neutralisers.

Impurities control:

− reducing chlorides, sulphurs and carbonates in order to obtain a conductivity belowto 0,2 microS/cm (piping)

− control of Na/Cl rate (<0,7) (steam generator)− avoid resins entrance (internals).

Methods of conservation and polishing during outage:

− maintaining the systems drained and dries or using wet lay-up (steam generator)− sludge and deposits polishing, using chemical and mechanical methods (sludge

lancing) or evaporation (feed-and-bleed).− control of hideout return and pollutant particles in the water-steam circuit on start-

up.− optimisation of purge systems− polishing of component within refuelling, avoiding contact with boric acid

Establishment of chemical guidelines to supervising and control

Addition of chlorides and biocides to control and reduce the microbiologicaldevelopment (open circuit).

57

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The modifications of operating conditions is other method used to mitigate degradationmechanisms as fatigue, corrosion, neutron embrittlement, as well as thermal ageing ofelectrical equipment. Following there is a brief explanation of the most commonmethods uses:

Reduction of transients number and severity

In order to reduce the incidence of thermal fatigue, it is a common practice to reducethe pressure and temperature transients affecting the component, as well as avoidstratification problems.

Some critical components where this practice could be used are: reactor pressurevessel, steam generator (nozzle and shells), internals and piping. For instance, in thecase of the RPV the operating conditions are controlled to limit the temperaturesdifferences to 200ºF (110°C) during heat-ups and cool-downs.

Temperature decreasing

Temperature is one of the factors affecting the corrosion process or thermal ageing ofseveral components, as steam generator, RPV and internals.

In the steam generator, a reduction of the temperature implies a decreasing in thereaction velocity and the evaporation rate, reducing at the same time the saltsconcentration and corrosive substances, and therefore the risks of different types ofcorrosion: IGSCC, pitting, etc. Nevertheless, its use is reduced to critical situations ,because of efficiency decrease. In this case, a technical-economical study isrecommended, in order to select the best cost-effective option.

In RPV and internals, a temperature control is a key factor to avoid or reduce thermalageing and corrosion. A reduction of the temperature in the heat branch from 315ºC to290ºC has provide optimum results to avoid SCC. Related to thermal ageing, thereduction of temperature normally implies studies to change the distribution of flowsand temperatures, to avoid local heating.

It shall nevertheless be mentioned that a reduction of system temperature increases thesusceptibility to neutron irradiation of the RPV.

Reduction of neutron flux

Degradation mechanisms as irradiation embrittlement or corrosion assisted byirradiation could be reduced at the vessel beltline and internals (core shroud) by meansa reduction of neutron flux by fuel element management (low leakage core or usingdummy fuel elements)

Reduction of normal operating current

This method could be used to mitigate the effects of thermal ageing in cables.

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This methods does not produce a real mitigation of degradation mechanisms itself, butsuppose an effective help in order to decide the most appropriated strategy to reduce itseffects, by means a component ageing evaluation.

For instance, operating conditions monitoring, specially temperature and pressuretransients, could provide fatigue usage factor, and therefore determine critical areasand conditions that implies a degradation risk and should be avoided if possible.Similar objectives follow other types of systems focussed on monitoring chemicalparameters, vibrations, etc.

58

In some component, for instance turbines, control systems limiting the heat-up velocityhave been installed, in order to avoid critical start-ups.

59

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Corrosion Resistant Cladding(CRC)

Placing of metal filler coating on the internalsurface of the piping adjacent to the weldflange to create a protective coating betweenthe coolant and the sensitised area of thematerial.

Carbide Solution Treatment(SHT)

Heating up of the sensitised area andsubsequent quick cooling to retain carbon.

Heat Sink Welding (HSW) isapplied to repair or replace asection

Minimises material sensitisation and createscompression stresses in the weld.

Last Pass Heat Sink Welding(LPSHW) is applied to repair orreplace a section

Identical to HSW where water is circulatedinside the weld during the last welding pass.

Weld Overlay StressImprovement (WOSI)

An extra weld bead is applied with cold waterrunning inside it.

Induction Heating StressImprovement (IHSI)

Similar to HSW where heating is by means ofcoils outside the piping while cold water runsinside it.

Change/Recovery ofMaterial Characteristics

Mechanical StressImprovements Process (MSIP)

Produces compression stress in the sensitisedarea by means of a hydraulic jack.

Piping IGSCC

Changes of OperatingParameters

Hydrogen Water Chemistry Injection of hydrogen to reduce the oxygencontent to 10 to 20 ppb

60

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Thermal Embrittlement Change/Recovery ofMaterial Characteristics

Annealing Annealing at 550ºC for 1 hour. The processhas hysteresis and is limited by the size andshape of the component.

Fatigue Changes of OperatingParameters

• Reduction of temperaturegradients

• Redesign of layoutpreventing flow stratifications

• Redesign of supportminimising flow vibrations.


Adjustment of water chemistryto values that minimise thepossibility of E/C

This method is limited by the cost-benefitevaluation of its installation and therequirements of the system itself.


Use of more resistant coating,or improvement of materialchemical composition

• Application of stainless steel coatings bymeans of welding

• Painting of service water systems• Minimum chromium content specified for

ferritic piping

Erosion/Corrosion

Change of SSC Design Redesign of the layout avoidingareas of high turbulence.

Corrosion, MIC Changes of OperatingParameters

Lay-up conservation methods • Wet conservation• Dry conservation• Inerting with nitrogen

Piping

Galvanic Corrosion Change of SSC Design Cathodic Protection

61


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Denting Changes of OperatingParameters

Reduction of corrosionproducts

Mechanical or chemical cleaning of the tubes,optimisation of blowdown systems, use ofcorrosion inhibitors (phosphates or AVT),inverse flow control, elimination of coppercomponents, control of NaCl ratio.


Use of All Volatile Treatment(AVT)

Wastage/Thinning


Cleaning


Reduction of the temperature ofthe primary at the inlet of thesteam generators

IGSCC/IGA/PWSCC


Stress relaxation Heat treatments or shot blasting.


Installation of sleevesPitting


Temperature reduction

Change of SSC Design Structural changes in tubesupports

Reduction and change of gaps between tubesand the AVB.

Fretting


Reduction of vibrations Optimisation of flows to minimise vibrations,installation of baffle plates

Fatigue Change of SSC Design Reduction of vibrations Optimisation of flows to minimise vibrations,installation of baffle plates

Fatigue (nozzles) Changes of OperatingParameters

Reduction of transients Minimisation of load changes and otherpressure and temperature transients

Steam Generator

Fouling Changes of OperatingParameters

Cleaning Flashing, chemical cleaning

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In service AnnealingNeutron Embrittlement


Neutron Flux Management


Temperature reductionIGSCC ( INCONEL600)


Surfacing of cracksPreventive Treatment of surfaces

Fretting Changes of OperatingParameters

Rotate, shift or exchangecomponents: thimbles, bolts, etc.

Reactor PressureVessel


Reduction of real loads Minimise pressure and temperaturetransients

Change of SSC Design Modification of circulation to coolhot areas

Change from down flow to up flowIGSCC


Adjustment of chemicalparameters to minimise theaggressiveness of the medium


Reduction of real loads Minimise pressure and temperaturetransients

Stress relaxation Changes of OperatingParameters

Modification of preloading tocompensate relaxation

Thermal ageing Change of SSC Design Redistribution of cooling flow

PWR Internals

Thermal ageing Changes of OperatingParameters

Exchange components

IGSCC Changes of OperatingParameters

Hydrogen Water Chemistry Hydrogen injection to reduce oxygencontent to 10 a 20 ppb

BWR Internals

IASCC Changes of OperatingParameters

Neutron Flux Management Reduce flow on the core shroud.

63


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Cracking Change/Recovery ofMaterial Characteristics

• Injection of epoxy resins• Shaping and sealing• Stapling• Boring and Covering• Flexible sealing• Slurry injection• Polymer soaking

Scalling/Spalling Change/Recovery ofMaterial Characteristics

Application of cement, epoxy resinor high quality concrete

Concrete Structure

Porosity Change/Recovery ofMaterial Characteristics

Injection of epoxy resins, slurry orepoxy foam


Control and renewal of coating(painting)

Metal Containment Corrosion

Change of SSC Design Cathodic ProtectionChange/Recovery ofMaterial Characteristics

Application of coating Painting on the outside and metallic orrubber lining in the inside


Use of corrosion inhibitors influids

Addition of sulphates in iron. It admitsmore inhibiting spectra in the closed cycle


Cleaning

Corrosion

Change of SSC Design Cathodic Protection Impressed current, sacrificial andprotective anode


Use of grafts and coating Limited use of metallic materials

Heat Exchangers andTanks

Erosion


Recasing

64


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Motors Thermal Ageing Change of SSC Design Optimisation of cooling systemsChargers and Inverters Thermal Ageing Change of SSC Design Optimisation of cooling systems

Change of SSC Design Installation of pressure transient dampers(dischargers, windings) at the equipment inletReduction of the normal operating current

Change of SSC Design Modification of the cable layout in trays inorder to optimise cooling

Cables Thermal Ageing

Change of SSC Design Placing cables from local heat sourcesMCC and Load

CentersThermal Ageing Change of SSC Design Optimisation of ventilation systems

65

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The safe management of NPP ageing is primarily dependent on managing the physical ageing ofsystems, structures and components (SSC). Basically, this calls for in-depth knowledge of theprocesses of degradation (particularly the kinetics and probability), the availability of on-site testequipment with which such deterioration can be diagnosed and methods and (material and human)means intended for repairing or replacing the parts whose serviceability has been impaired by thedegradation.

This also implies sustaining an industrial fabric that has the means of providing spare parts to securereplacement or repair of degraded components.

But in addition to the ageing-induced material changes in the SSC features, other changes are ofequal significance (particularly during regular safety reassessments):

− developments in techniques and documents since the creation of the power plant (changingdesign methods and computer codes, non-destructive testing methods, industrial codes andstandards...),

− developments in technology/obsolete systems (example of analogue systems being replaced bydigital systems),

− developments in safety regulations and standards and acceptable dose limits for staff in terms ofradiation protection.

− developments in information relating to accidents caused by an external source (earthquake,flooding ... ) or an internal fault in the design or use.

Further broader changes, associated with the industrial sector or society, are being investigated:

− safeguarding of skills and knowledge following the retirement of a number of experts who hadbeen involved in the design, construction, commissioning and operation of power stations,

− training of new staff in the nuclear industry (experts and also non-specialised staff and completeplant suppliers),

− preservation of the scientific infrastructure (particularly R&D services) within a nuclear-hostileenvironment,

− developments relating to safety culture over long periods, with or without staff turnover,

− adaptation of organisation models, work methods (global industrial transformation). In additionto current technical management which has already improved the safety of plant units (throughupgrading or modification of SSC), industrial management will also contribute factors ofprogress (ergonomics, streamlined organisations, greater emphasis on the human factor...).

Sharing the feedback from these different areas with the various partners in the European nuclearindustry should also contribute to strengthening and intensifying management of ageing and changein general.

66

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Obsolescence is a non-physical mechanism that could affect the capability of components, systems,or structures to perform their required functions. Obsolescence means the potential unavailability ofspares due to the evolution of the market or/and the technical progress.

Obsolescence is not only limited to SSC but is also applied to the knowledge and the know-how.This risk of loss of required knowledge and know-how could also affect the level of safety andefficiency of the NPP.

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After conception of the foetus - or after impairment due to an accident or disease - the number ofcells in the human organism capable of self-regeneration is relatively limited. Typical continuouslyregenerating cells are the liver cells, blood cells and bone marrow cells; other examples include thescarring of injuries and the renewal of the skin, nails and hair. The human organism as a whole issubject to a still very poorly understood ageing process which degrades the performance of manyhuman functions. In fact, the genetic code determines the life span and cell regenerability. Certaincells regress very early on, as for example, the nerve cells in the corpus callosum1, whose numberdrops significantly at the end of pregnancy, i.e. even before birth.

The human organism attains its optimum development at the age of 20-25 years though the initialeffects of ageing begin even before this time as the sensory receptors start to lose their sensitivity.The effects on the visual and auditory cells are however better known thanks to the systematicmedical follow-up practised. In the forties (40-45), the eye’s crystalline lens starts to harden andgives rise to presbyopia. As far as the loss in auditory sensorial sensitivity is concerned, thedifferent tones are not affected in the same way: certain syllables and certain words are morefrequently distorted than others. The human hearing system is also less able to synchronise soundstimuli in the two auditory tubes. Laboratory experiments have demonstrated that signals sentseparately to the two ears are not always properly synchronised compared to the signal transmissionsequence: this affects the ability to localise the exact direction of sound sources.

From the age of 20-25, the short-term memory performance begins to regress very slightly. Theworking memory is less flexible which reduces intellectual agility. The long-term memory has aharder time encoding the information to be memorised, which has to be manipulated many moretimes before it is sufficiently salient and can located and extracted as easily as before. There aremore "objects manipulated by the reasoning process" due to the greater amount of experience builtup. Overall inferential reasoning becomes more efficient, more dependable and has a wider field ofapplication. With age, the brain works better by intuition than when young.

Nonetheless it should be noted that the memory development cycle starts from a very early age andis marked by phases with different mnestic capacities. In reality, it is more correct to speak of acontinual evolution of the memory throughout the life of a person rather than ageing.

Taken together, the different effects of ageing change the ability to study. This first becomesnoticeable at about 25 and, after 30, the adult finds it hard to follow the pace of learning, perfectlysuitable for students ten years younger.

The decrease in learning capacity also seems related to the lack of practice in this kind of activity.With increasing age the learning ability - in scholastic terms - lessens; this should be taken intoaccount in view of the increasing requirements for retraining and the development of training in thecareer cycle.

1: The corpus callosurn is an arch connecting the two cerebral hemispheres and consisting of neurons.

67

As intellectual and sensorimotor agility decreases, certain types of knowledge acquisition becomemore laborious: one example is the use of computers, which simultaneously calls for learningkeyboard data entry and assimilating a new thought process to match the logic and themanipulations required respectively by the commands and the software.

The quality of sleep also lessens with age: the sleeping pattern is more fragmentary and the sleeperis more sensitive to noise and the surroundings. Falling asleep and going back to sleep are harderand take longer.

Ageing affects the regulation of the metabolism: at the age of 50, the basal metabolism hasdecreased by about 5% and at 60 by 10%. This leads to an overall drop in the mobilisation of theorganism’s internal energy resources and an ensuing reduction in physical powers. Recovery isslower due to catabiosis (cell ageing) and the diminished renewal benefits of sleep; the individual’sresponsiveness and adaptability also regress. Moreover ageing influences the capacity to indulge inintense effort, to keep up with high work rates and to adapt to difficult conditions, such as nightwork. A study of female workers on a high-productivity assembly line detected this phenomenon asearly as the age of 25.

Thus the "coping response" acts, as it were, in the background. Despite the lower physiologicalperformance, the human being regulates overall his activities better and allows for his capacitiesand reactions. A strategy of anticipation is adopted and work activities are organised differently bystriving to optimise physical exertion. His life is organised differently by making more effective useof his power of recovery.

As Welford said in 1973: "A full understanding of the task in question partly or entirelycompensates for the effects of ageing".

Mental processes are also subject to the same kind of adaptation. The lower effectiveness of theshort-term and the working memories reduces the direct mental capacity for analysis. However thisimpairment is largely compensated for by a greater mastery in handling inferences and by a moreextensive experience (including potential error sources). Logical algorithmic reasoning is moreproblematic and intuitive capacities expand, backed by a long experience. Diagnosis and heuristicreasoning processes are more complex, more cross-linked with more iterations, the checks moredeveloped and the solutions considered in greater detail as to their effects.

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The operation life-time is the common factor contributing to the ageing process of components andstructures and additionally it provokes a relevant impact in terms of the staff ageing and the need ofretirements and substitutions by less experienced people.

• Most nuclear plant workers have left the industry during the past years, mainly retired but alsobecause they have been hired by more promising industrial sectors. People, which are going tobe retired during the next years, are an important aspect usually not considered in NPPs ageingplans.

Criticality of this point increases if it is considered that, at the same time, the ageing plant willhave need of higher efforts in equipment surveillance and maintenance.

• A different problem is time relaxation of worker habits and behaviours. Work practice trends toreduce the adaptation capability to the new situations: new instruments, uprating and designmodifications, which could require different operators training, etc.

Theoretically the above-mentioned problems are covered by the requirements imposed by theRegulatory Authority concerning the qualification and training of the NPPs operators. In Spain,specific Safety Guides (GS-05.6 [100] and GS-07.04 [101]) issued by the Spanish NuclearRegulatory Authority (Nuclear Safety Council, CSN) define the requirements for personnelselection, specific training adapted to the plant singularities and periodic re-evaluation of the

68

capabilities and skills of the staff members. However, preservation of the knowledge related toageing, their contributing factors and their evolution, as well as the mitigation, control andmonitoring methods and techniques should be part of the new staff training.

To avoid the problems, the following actions could be performed:

• Periodical training for personnel requalification.

• Rotations of working places to reduce stresses and increase workers interest.

• Take on new staff in time, ensuring an overlap between new workers and elder ones.

• Use retired workers as a source of experience and advice. As example, some plants are usingthem during outages or to help with specific problems.

• Recruit expert staff from other industry, which could help to integrate new knowledge in thenuclear sector.

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Oldest plants need more attentive behaviour of staff in order to maintain operating performance.Competitiveness, safety levels and availability of plants are associated to plant staff culture. Someimportant ageing related aspects, among others, are:

• Decisions and actions of plant staff considers plant safety as a priority when plant ages:

− When conditions are outside procedures and policies, use of conservative approaches.− Operational decisions based on safety analysis.

• Behaviours that contribute to excellence in human performance need to be reinforced:

− Strengthening of communications.− Anticipate problems.− Eliminate conditions that lead to human error.− Correct procedure deficiencies, if any.

• A line organisation is effectively implemented and maintained:

− Improvement in administrative controls, policies, procedures and schedules for activitiesaffecting safe and reliable plant operation.

− Design, manage and improve key processes to contribute to safe and reliable plant operation.− Organisation preparing utility personnel to mitigate the consequences of core damage and

manage emergency situations.

• Use training to achieve, improve and maintain high level of personnel knowledge skill andperformance:

− Improve personnel performance.− Maintain and improve job-related knowledge and skills.− Identify similar precursor conditions and initiating proactive corrective actions− Continuous training for periodic reviews of applied fundamentals.

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In this context, the Spanish NPPs have structured their Life Management Programmes in order toprovide the information required performing the continuous condition evaluation. This informationprovides a powerful technical support to the staff in the ageing control field.

The afore-mentioned information is obtained, among others, in the following sources:

• In service inspections, examinations and tests as required by the current Licensing Basis.

69

• Data and information obtained during maintenance tasks. The maintenance evaluation andimprovement for life management programmes have identified the required information.

• Specific ageing monitoring and condition evaluation systems.

• Analysis of the impact of operating experiences findings.

• New regulatory requirements, notices and bulletins generated by the finding of new technicalaspects on ageing.

A well structured and updated information covering all the facets of the significant degradations isthe best guarantee to support a precise condition evaluation keeping the historical data available forthe NPPs staff along their service life.

70

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�� %(/*,$1�$3352$&+In Belgium, there is no predetermined lifetime for a nuclear power plant, either license life or eitherdesign life. Belgian nuclear power plants may be kept operational as long as they can operate safely(i.e. to maintain the operating license). Economical aspects will also define whether or not theUtility will continue operation.

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Seven nuclear reactors are operating in Belgium. They are located at two sites and have a total netcapacity of about 5.7 GWe. All units are equipped with Pressurised Water Reactors (PWRs). Table4.1 presents an overview of the reactors, their power and the year of commissioning.

The Belgian producers (Electrabel and SPE) have a share (capacity reservation) of 25% of the B1and B2 units of the French Chooz NPP.

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Most of the regulations concerning nuclear installations in Belgium are contained in the RoyalDecree of February 1963 for the Protection of the Population and of Workers and its subsequentupdates (last update: October 1997).

Operating licences for nuclear installations are granted by Royal Decree, countersigned by theMinisters of Interior Affairs, on the advice of the National Special Commission for Nuclear Safety.

Two Government departments: the Radiation Protection Service and the Technical Safety Servicefor Nuclear Installations are responsible for the implementation and control of the regulations aswell as the operating licences. Since the law of April 1994, the Federal Agency for Nuclear Controlhas been created and the two Government departments were incorporated in this organisation. ThisAgency is responsible for the inspection and surveillance of nuclear activities in Belgium.

All permanent control tasks and monitoring of the activities of operators are performed by theAuthorised inspection organisation (Association Vinçotte Nuclear, AVN), licensed for this purposeby the Government.

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71

All Belgian nuclear power plants have no time-limited authorisation of operation; hence there is nopredetermined license life. The safety behaviour is assessed on a continuous basis by the Safetyauthorities and, as required in the license of each nuclear power plant, an overall safety review mustbe performed every ten years. The list of the subjects to study during the safety review isestablished on a common agreement between Association Vinçotte Nuclear and Electrabel /Tractebel Energy Engineering. The oldest plants (i.e. Doel 1&2 and Tihange 1) have gone throughthis process twice while the last four units (i.e. Doel 3&4 and Tihange 2&3) only once.

The safety reviews, initiated by the Utility, must compare on the one hand the condition of theinstallations and the implementation of the procedures that apply to them, and on the other hand theregulations, codes, and practices in force in the United States and in the European Union. Thedifferences found must be highlighted, together with the necessity and possibility of remedial actionand, as the case may be, the improvements that can be made and the time-schedule for theirimplementation. This examination is dressed in a synthesis report identifying differences observedand assessing the necessity and possibility to put them right; plant modifications and proceduresimprovements are planned. The identified subjects are then arranged according to the structure ofthe Safety Analysis Report. For the next safety review, the subjects might be treated according tothe IAEA Safety Guide N° 50-SG-012 "Periodic Safety Review of Operational Nuclear PowerPlants" [50].

The main objectives of the safety review are to ensure that the plant is still as safe as it was when itsoperating licence was granted; to take account of any possible future deterioration of equipment inorder to ensure safety for the ten years to come; and to improve general safety by making anychanges thought reasonable or necessary in the light of the most recent safety standards andpractices.

A systematic analysis of experience feedback from the Belgian plants and plants abroad results inimprovements to systems and/or replacement of components, verification of the coherence of pastmodifications. Ageing is thus systematically investigated in order to demonstrate that the safety ofthe installations is guaranteed during the next decade.

To minimise the economic impact, the replacements of large components (e.g. steam generatorreplacement) are usually executed at the same time as the implementation of the main changesresulting from these ten-yearly safety reviews.

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Again, there is no predetermined design life for the entire nuclear power plant. Design life is merelyused in the design specification of some components subjected to known degradation processes.Design life includes values such as 20, 30 and 40 years. The first value was taken into account inthe first Belgian-French commercial nuclear power plant of Chooz A for the design of somecomponents. In the more recent plants, component design life is 40 years.

Component design life can be modified by re-qualification (e.g. fatigue analysis more detailed thanthat of the design) or by considering the real values of the parameters affecting the degradationprocess instead of the hypothetical design values (e.g. real transients, real number of occurrences oftransients for components subjected to fatigue).

Examples of components with a predetermined design life are:

− Components subjected to fatigue due to low cycle thermal and pressure transients, such asprimary components and piping. The number of occurrences considered in the original designfor those components corresponds to 40 years, except for Doel 1&2 and Tihange 1, where 30years was originally considered. For Tihange 1 the power upgrading studies, during which thefatigue analyses were re-evaluated, have brought up this duration to 40 years for the majorprimary components;

− Components subjected to irradiation embrittlement, such as the reactor pressure vessel;

72

− Primary containment tendons subjected to pre-stress relaxation; and

− Electrical and mechanical equipment subjected to severe environment conditions and having apredefined “qualified life”. Those qualified lives have duration which are component dependent.

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The government, which is responsible for the safety of persons and property, sets objectives andacceptability limits as regards nuclear safety; the ministers for the environment and industry arecurrently jointly responsible for the technical regulations in this field.

The requirements and prescriptions of the general technical regulations covering the nuclear safetyof pressurised water reactors are formulated in two ministerial orders: that of 26 February 1974specifies the way in which pressure vessel regulations are applied to the main primary systems ofnuclear steam supply systems and that of 10 August 1984 deals with the design quality,construction and operation of major nuclear installations.

As far as plant operation is concerned, the 1974 Order has been completed by the November 10,1999 order, which also applies on the Main Secondary System of NSSS.

Further to the review of the safety options open to nuclear units in the most recent standardisedseries, ministerial directives published in 1979 and 1983 established the obligations and principalfeatures of installations in the 1300 and 1400 MWe series, particularly as regards safety.

Certain letters and statements of position emanating from the Nuclear Installations SafetyDirectorate (DSIN) establish jurisprudence in the field of nuclear safety in France.

Furthermore, the basic safety rules laid down by the Nuclear Installations Safety Directorate and itstechnical support bodies specify the conditions which have to be met if operation is to be consideredas adequate and complying with regulatory practice.

The Design and Construction Rules, which are codes and standards established by the Frenchnuclear industry, express what is considered as good professional practice adapted to the nuclearindustry, meeting the obligations associated with licences; likewise, there are Rules for Monitoringduring Operation. These rules are periodically discussed with the Safety Authority and theiracceptability is recognised through the issuing by DSIN of a Basic Safety Rule (RFS) dedicated tothe use of each approved code. Codes are then accepted, but not mandatory.

To support their applications for construction, start-up, commissioning and subsequentdecommissioning licences, nuclear operators have to produce Safety Analysis Reports and GeneralOperating Rules; guidance as to how these should be drafted and reviewed (safety analysis) is givenin the ministerial instruction of 27 March 1973.

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The Nuclear Installations Safety Directorate ensures that the level of safety of each unit ismaintained despite the effects of ageing and that the inevitable shutdown of an installation isprepared sufficiently well in advance.

The regulatory provisions concerning the lifetime of units are, however, fairly restricted at present;there is no limit as to how long units can go on operating but nuclear operators must be able todemonstrate the safety of their facilities whenever required to do so.

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A periodic review process exists but is not required by law.

73

The decree of 11 December 1963 on major nuclear installations was modified on 19 January 1990by the introduction of a clause stating that the ministers of industry and the environment were freeto require, conjointly at any time, that nuclear operators review the safety of their installations.

In practice, the Safety Authority requires that nuclear power plant safety be reviewed every tenyears, supplementing the safety analysis conclusions drawn from experience feedback from 58relatively recent plants (at 1 January 2000, the 900 MW plants had an average age of 17 years andthe 1300 MW plants an average age of 11 years: Table 4.2) and based on detailed technical dialoguewith the single nuclear operator.

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Chooz A 320 PWR 4 – 1967 24(Plant decommissioned

end of Nov. 1991)Belleville 1 1363 PWR 10 – 1987 13Belleville 2 1363 PWR 7 – 1988 12Blayais 1 951 PWR 6 – 1981 19Blayais 2 951 PWR 7 – 1982 18Blayais 3 951 PWR 8 – 1983 17Blayais 4 951 PWR 5 – 1983 17Bugey 2 945 PWR 5 – 1978 22Bugey 3 945 PWR 9 – 1978 22Bugey 4 917 PWR 3 – 1979 21Bugey 5 917 PWR 7 – 1979 21Cattenom 1 1362 PWR 11 – 1986 14Cattenom 2 1362 PWR 9 – 1987 13Cattenom 3 1362 PWR 7 – 1990 10Cattenom 4 1362 PWR 5 – 1991 9Chinon B1 954 PWR 11 – 1982 18Chinon B2 954 PWR 11 – 1983 17Chinon B3 954 PWR 10 – 1986 14Chinon B4 954 PWR 11 – 1987 13Chooz B1 1516 PWR 8 – 1996 4Chooz B2 1516 PWR 4 – 1997 3Civaux 1 1516 PWR 12 – 1997 3Civaux 2 1516 PWR 12 – 1999 1Cruas-Meysse-1 956 PWR 4 – 1983 17Cruas-Meysse-3 956 PWR 5 – 1984 16Cruas-Meysse-2 956 PWR 9 – 1984 16Cruas-Meysse-4 956 PWR 10 – 1984 16Dampierre-1 937 PWR 3 – 1980 20Dampierre-2 937 PWR 12 – 1980 20Dampierre-3 937 PWR 1 – 1981 20Dampierre-4 937 PWR 8 – 1981 19Fessenheim 1 920 PWR 4 – 1977 23Fessenheim 2 920 PWR 10 – 1977 23Flamanville-1 1382 PWR 12 – 1985 15Flamanville-2 1382 PWR 7 – 1986 14Golfech-1 1363 PWR 6 – 1990 10

74

Golfech-2 1363 PWR 6 – 1993 7Gravelines B-1 951 PWR 3 – 1980 20Gravelines B-2 951 PWR 8 – 1980 20Gravelines B-3 951 PWR 12 – 1980 20Gravelines B-4 951 PWR 6 – 1981 19Gravelines C-5 951 PWR 8 – 1984 16Gravelines C-6 951 PWR 8 – 1985 15Nogent-1 1363 PWR 10 – 1987 13Nogent-2 1363 PWR 12 – 1988 12Paluel-1 1382 PWR 6 – 1984 16Paluel-2 1382 PWR 9 – 1984 16Paluel-3 1382 PWR 9 – 1985 15Paluel-4 1382 PWR 4 – 1986 14Penly-1 1382 PWR 5 – 1990 10Penly-2 1382 PWR 2 – 1992 9Saint Alban-1 1381 PWR 8 – 1985 15Saint Alban-2 1381 PWR 7 – 1986 14St Laurent B-1 956 PWR 1 – 1981 20St Laurent B-2 956 PWR 6 – 1981 19Tricastin-1 955 PWR 5 – 1980 20Tricastin-2 955 PWR 8 – 1980 20Tricastin-3 955 PWR 2 – 1981 20Tricastin-4 955 PWR 6 – 1981 19

This review compares the level of safety of an installation with:

− its original level of safety, making sure that the requirements of the original operating licencecontinue to be met through the identification of any damage and the review of weak pointsrequiring further analysis or justification (review of conformity),

− the safety of the most recent plants in operation, the aim being to upgrade the safety of unitsbeyond the initial design level in the light of progress made due to increased knowledge in fieldsfor which little experience feedback is available and continuous analysis of operating experience(safety reviews).

These regular reviews of plant series and the corresponding 10-yearly inspections of units shouldgive the nuclear operator the opportunity to carry out in-depth reviews which are more exhaustiveand far-reaching than those initially planned in basic preventive maintenance programmes.

During these prolonged outages, the nuclear operator carries out, where necessary, any time-consuming repair work or replaces large items of equipment, assuming that this was not requiredurgently beforehand.

But in the eyes of the Safety Authority, this provides no more than a snapshot of the state of thecomponents, systems, equipment and structures which contribute to the safety of the installation.

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Since the nuclear operator must be in a position to demonstrate the safety of its installation at alltimes, the approach whereby safety and extended operating life are linked should be supplementedby a demonstration that the effects of ageing are under control until the next inspection is made, aswell as during the inspection made at the time in question.

When demonstrating that the effects of ageing are under control, the nuclear operator should belooking to the future as regards all aspects of its installation; the demonstration should be thesubject of permanent technical dialogue with the nuclear operator and should be based on:

75

− the results of detailed experience feedback, organised horizontally throughout the plant,

− increased knowledge of ageing phenomena, their origins and the way they develop and thekinetics of deterioration, through R&D work; progress would appear to be possible, particularlyas regards estimating loads and identifying sensitive areas,

− the development of qualified methods for decontaminating lines and equipment in preparationfor inspection, repair and replacement operations,

− the implementation of non-destructive methods for inspecting components and structures andassessing materials, be they already available or still to be developed and qualified,

− the identification and monitoring of ageing indicators and criteria applicable to the plant,

− justification files and inspection reports for assessing whether equipment should be left inservice or treated, modified, repaired and/or replaced,

− files indicating the technical and organisational feasibility of exceptional maintenanceoperations.

Experience feedback has revealed that the first two preventive lines of defence, namely makingallowance for ageing at the design stage and monitoring ageing, may fall short; the nuclear operatorcould then find itself torn between allowing operation to continue in degraded conditions orcondoning outage for an indefinite period.

The insistence of the Nuclear Installations Safety Directorate on the need to look sufficiently farahead when envisaging possible repair, replacement or modification operations stems from the riskof conflict between availability and safety which would result if there were no such approach andresults in a third line of defence.

This risk is to be taken all the more seriously since around 80% of electricity is generated in unitswhose design is sufficiently standardised for them to be liable to be affected by generic degradationassociated with ageing.

Extension of the operating life of each installation, for a given period, could be envisaged in thiscontext on the basis of examination of the results and conclusions of the highly-important third 10-yearly inspection.

Furthermore, as the units age, demonstrating that the effects of ageing are under control willincreasingly form part and parcel of the operation monitoring files and will gradually become adaily preoccupation in the plants.

It should, however, continue to constitute the best opportunity for bringing together the lessonslearned from experience gained in the plants, analysis carried out by corporate departments and theresults of research and development work.

Each unit should be treated on a case-by-case basis, making allowance for the construction,manufacturing, operating and environment-related differences between reactors in the same series.

Life management of units should take these special features into account but can only be undertakenin the context of generic problems considering the population of plants as a whole.

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In Spain, as in most European Union countries, the NPP operating license is open regarding itsduration, there being no legal restrictions for extending the operational life of the NPPs by renewingtheir licenses.

Spanish NPPs owners and Nuclear Regulatory Authority (CSN, Consejo de Seguridad Nuclear –Nuclear Safety Council) are interested in realising an effective management of components ageing

76

processes. Therefore, CSN has introduced within its inspection and control functions specificrequirements related to ageing management.

Nowadays, the regulatory requirements related to NPP Lifetime Management are basically thefollowing:

- NPP operation license requires preparing and submitting to the CSN an annually updated reporton the Ageing Control Activities or the Lifetime Management Programme.

- Continuous NPP safety evaluation process by CSN is complemented with Periodic SafetyReviews (PSR), to be performed every 10 years, including: a) review of components behaviour(identify degradation mechanisms and current corrective measures adopted by the plant forageing mechanisms control and mitigation) and b) updating of safety evaluation andimprovement programmes (Lifetime Management Programme is included among them).

The following paragraphs describe in more detail the safety and regulatory aspects applicable toSpanish NPPs, which are listed in Table 4.3.

77

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José Cabrera (Zorita) 160 PWR August 69 31Santa Mª de Garoña 466 BWR May 71 29Almaraz 1 974 PWR September 83 17Almaraz 2 983 PWR July 84 16Ascó 1 979 PWR December 84 16Cofrentes 1,025 BWR March 85 15Ascó 2 1,014 PWR March 86 14Vandellós II 1,081 PWR March 88 12Trillo 1,066 PWR August 88 12

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Nowadays, the regulatory requirements related to NPP Lifetime Management are the following:

• Currently, NPP operation license require to prepare and submit to the CSN an annual reportincluding a Lifetime Management Programme.

• Nuclear Regulatory Authority continuous safety revisions are complemented with PeriodicSafety Reviews (PSR), performed every 10 years, including among other aspects:

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- Identify the ageing or degradation mechanisms

- Submit the corrective measures adopted by the plant for ageing mechanisms control andmitigation.

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Every year, NPPs report to CSN about Lifetime Management Programme status as well as activitiesperformed within this Programme.

In this way, in February 1998, a generic methodology, developed by the Electricity Sector(UNESA) was audited and accepted by CSN. This methodology covers the main aspects consideredby the guide of IAEA-TRS-338 "Methodology for the Management of Ageing of Nuclear PowerPlant Components Important to Safety" [37], and considers:

• Component selection and grouping• Degradation mechanisms affecting components• Evaluation of Maintenance practices.

The Spanish NPP are following the above-mentioned methodology in order to perform theirspecific Lifetime Management Programmes. In fact, the Spanish NPPs have developed andsubmitted to the CSN their specific Lifetime Management Programmes and/or ageing relatedactivities from 1998 through 2000. These programmes have been audited and evaluated by the CSNin the frame of the CSN Safety Guide 1.10 point 4.4 "Status Updated of the Safety Evaluation andImprovement Programmes - Lifetime Management Programme".

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78

Periodic Safety Reviews requirements are explained in a specific guide (CSN Safety Guide N° 1.10[110]) published by the Spanish Nuclear Regulatory Authority in December, 1995. The contentsincluded in this guide are explained in more detail in the following sub-sections.

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During NPP operation, the CSN realises a continuous safety revision, through its inspection andcontrol function, and evaluation the following required analysis:• New codes and standards applicability• Operating experience analysis• Safety analysis of design modifications

The experience shows that these periodical reviews could be complemented considering newpractices, ageing cumulative effects and new modifications performed, as well as new technologiesdeveloped. Therefore, the PSR have been introduced, with the objective to perform an integralsafety review, taking into account different aspects as operation, maintenance programmes,equipment qualification and ageing management, in-service inspection results, codes changes andthe results applying the Probabilistic Safety Analysis (PSA).

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The PSR does not replace the normal analysis, control and surveillance practices, but realise eachten years a global evaluation of NPP safety and the potential improvements to be introduced,considering the components and structures status, as well as the new practices and codes applied inmodern NPPs.

The main objectives of the PSR are:

1. Guarantee the correct application of operating experience analysis process, including themodifications global revision.

2. Analyse the plant behaviour during a long operating period, including the results of surveillancerequirements and equipment maintenance, in order to verify the safety level has not decreasedand to guarantee the safety operation during the next period.

3. Plant Safety level evaluation, considering the new national codes and internationalrecommendations, in particular those applied in the project origin country to the similar NPPs.

4. Update the safety evaluation and improvement programmes.

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The PSR main criteria are:

1. Use of the project origin country codes and laws, complemented with internationalrecommendations and new codes, when applicable.

2. Considers the current plant license basis.

3. The internal operating experience analysis will use as reference those events notified accordingto Technical Specifications of the plant.

4. The external operating experience analysis will use as reference those significant events definedby recognised organisations, according to the technology of project origin country and mainsupplier technical bulletins.

5. The Probabilistic Risk Assessment (PSA) is an important tool to analyse plant Safety, andtherefore will be an important issue in the PSR.

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The following four aspects have to be analysed in the PSR:

79

1. Operating experience

The plant should demonstrate that it reviews continuously the operating data and maintain a highsafety level with a feedback of operating experience, detecting and analysing and correctingadequately every potential dangerous conditions.

The NPP should perform:

• Revision of operating experience, including personnel doses, operating indicators, wastegenerated and environmental impact.

• Revision of operating experience of other similar NPPs.• Revision of suppliers communications, about deficiencies or faults detected in their equipment,

and the correctives actions applicable.• Revision of�NPP documentation, including design, manufacture, construction, testing,

maintenance, qualification, modifications, faults, incidents, etc.• Analysis of tendencies to demonstrate that its behaviour follows the design previsions and the

extrapolation of lifetime is acceptable.• Revision of�all corrective actions related to safety.• Revision of safety analysis performed related to design modifications to guarantee that the

design basis have not been affected.

2. Component behaviour analysis

The NPP will check that the safety equipment status surveillance is appropriate to detect and correctits ageing during the service life.

The following actions will be performed to verify the above mentioned:

• Analysis of regulated maintenance programmes results, including the application of themaintenance Rule or equivalent programme.

• Review of in-service inspections results, analysing the relations between the defects detectedand the operating conditions and to propose corrective actions.

• Review of results of Technical Specifications monitoring and periodical tests.• Review of equipment qualification status, including the qualified and dedicated spare parts

management.• Identification of degradation mechanisms and corrective actions performed of foreseen.

3. New safety codes and standards impact analysis

It should be checked that the national and project origin country regulations, as well as theinternational recommendations, have been properly analysed.

The analysis process is:

• Consider the following regulation:

− American project design NPPs: Title 10 of Code of Federal Regulations, NRC RegulatoryGuides, NRC Generic Letters and NRC bulletins.

− Trillo NPP: Recommendations of German Reactors Safety Commission (RSK),Recommendations of Radiological Protection Commission (SSK), Nuclear Safety TechnicalCodes (KTA) and American regulation included in the license basis,

− All NPPs: Recommendations of international organisations, and in particular the documentsof Nuclear Safety Codes (NUSS) of IAEA.

• Evaluate the detected deviations, producing a programme of corrective actions.

• Use the PSA to establish the importance of deviations and the benefit of corrective actions.

4. Updating of safety evaluation and improvement programmes

80

The PSR considers the on-going programmes, including the Probabilistic Risk Assessment, in orderto identify the measures adopted for safety improvement.

Among others, the programmes for safety evaluation and improvement considered in the CSN guideare:

• Management of severe accident• License and non-license personnel training• Organisation and human factors• Application of Safety Culture concept• Operating procedures, specifically Emergency operating procedures• Improvement of Technical Specifications• Lifetime management programmes• Quality assurance• Doses reductions• Control and surveillance of radioactive effluents• Environmental radiological protection• Management of radioactive wastes• Power plant configuration control• Evaluation plans and systematic independent revisions.

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Periodic Safety Review results have to be reported in a specific document produced by NPP owner,including the analysis of all areas under the scope of the PSR, and identifying the safetyimprovements actions and its implementation schedule.

Once the document has been reviewed and analysed by the CSN, a final report will be produced,including the conclusions and the programmes to be implemented.

This final report will be the base for:

• Guarantee the adequate safety level.• Following continuos and / or periodic safety reviews.• CSN and NPP Action Plans, in safety significant areas.

The CSN Safety Guide 1.10 [110] have been prepared following international practices andrecommendations, including IAEA Safety Guide "Periodic Safety Review of Operational NPP".

In order to facilitate the application of CSN Safety Guide 1.10 "Periodic Safety Reviews of NuclearPower Plants", UNESA has published a specific guide, titled «Guideline for the development ofPeriodic Safety Review" [111] that is explained in the following chapter.

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The objective of this document is to establish the detailed scope and contents of Periodic SafetyReviews, that the Spanish NPPs should submit to the CSN, according to the requirements of theOperating License, and which criteria, objectives, responsibilities and basic scope have beenincluded in the CSN Safety Guide 1. 10.

The structure of UNESA document is similar to the CSN Guide, explaining in detail each areacontaining in the PSR and providing different types of forms to facilitate the compilation ofrequired information.

The areas considered within the UNESA guideline are the following:

• Operating experience• Experience related to the radiological impact

81

• Changes in the regulations and laws• Equipment behaviour• Installations modifications• Probabilistic safety assessment• Updating of safety evaluation and improvement programmes

The two areas related to ageing and lifetime management are: equipment behaviour and updating ofsafety evaluation and improvement programmes. Therefore, the following chapters have beenfocussed in these two areas, providing a brief explanation of the information to be collected by theNPP and the contents to be included in the reports of the PSR to be submitted to the CSN.

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4.3.5.1.1. Revision period

In the case of the first PSR, the period has been defined from the beginning of plant operation untilthe date of this first PSR. For subsequent PSR, the period follows the general criteria, that means 10years.

4.3.5.1.2. Revision scope

The PSR scope will include: maintenance, in-service inspection, technical specificationssurveillance requirements, equipment qualification and lifetime management.

4.3.5.1.3. Information sources

0DLQWHQDQFH�5XOH�(MR): The information needed is the project documents used for MRimplementation and for monitoring the rule application.

,Q�6HUYLFH��QVSHFWLRQ : The necessities of information are: Last interval (10 years) evaluationreports, In-service inspection handbook, Reports of the In-service Inspection results, Regulatorybody (CSN) audits and evaluations of results reports, documents of exemption requested.

7HFKQLFDO�6SHFLILFDWLRQV�VXUYHLOODQFH�UHTXLUHPHQWV: The documents of the compliance with thetechnical specification requirements are controlled by QA departments, as well as by CSN audits.

(TXLSPHQW�TXDOLILFDWLRQ� 7KH�required information includes: List of equipment, which requiresenvironmental and seismic qualification, Final Safety Analysis Report (FSAR), Environmentalqualification reports/dossiers, Plan for environmental qualification maintenance, seismicqualification documents, and component dedication documents.

/LIHWLPH� PDQDJHPHQW� Documents of the four previous points, related to results and mitigationmeasurements, lifetime management preliminary activities, lifetime management programmes, aswell as the annual reports required by the provisional Operating license.

4.3.5.1.4. Report content

0DLQWHQDQFH�5XOH��The following data have to be included:

• In the first PSR report: Maintenance Rule implementation programme and current status of theMR implementation.

• In all PSR reports, the list of Structures, Systems and Components (SSC) includes in the scopeof the MR, indicating the modifications in this lists as well as risk significance SSC, thesurveyed SSC, the action taken and the SSC classification following 10CFR50.65 criteria [112].In addition, trends and analysis results and future foreseen actions.

,Q�6HUYLFH�,QVSHFWLRQ. Related to In-Service Inspection programme, the PSR report includes:

• Evaluations of the last finish interval (10 years).

82

• Summary of modifications to the In-Service Inspection handbook in relation to the previousinterval.

• Inspections performed in the current interval. List of reports generated.• Exemptions requested• ASME XI code cases• Relevant event analysis.• Results evaluation and future activities.• Global analysis of barrier behaviour.

7HFKQLFDO�6SHFLILFDWLRQV�VXUYHLOODQFH�UHTXLUHPHQWV� In�addition to the ISI programme, the PSRreports include:

• Significant results of Surveillance requirements• Approved exemptions• Results evaluations and trends analysis.

(TXLSPHQW�TXDOLILFDWLRQ��The PSR reports consider:

• List of equipment contained in the Environmental qualification programme.• Environmental qualifications report status.• Environmental qualification maintenance programme. Qualified life.• Seismic qualification in accordance with codes and applicable requirements.• Qualified spare parts management.• Component dedication programme.• Results evaluation and future activities.

/LIHWLPH�PDQDJHPHQW�The PSR reports includes:

• Identification of transient. The transients are counted, quantified and trends analysis areperformed in order to verify that plant behaviour is into design foreseen conditions and the plantlife extrapolation is acceptable.

• Ageing and degradation processes are identified and the mitigation measurements have beentaken.

• If a formal Lifetime Management Programme exists for the power plant, the following pointsare included:

- Lifetime Management Programme implementation status.- More significant activities performed in the frame of the Lifetime Management Programme.- Lifetime Management Programme future activities.

• If the plant have programmed preliminary activities, which could allow to perform a LifetimeManagement Programme, the report will inform about the evaluation of annual reports requiredin the provisional Operating license and the development of the above mentioned activities andtheir future programming.

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The NPP should inform the CSN about the status of programmes related to safety evaluation andimprovement. The Lifetime Management Programme has been included within these programmes.

4.3.5.2.1. Revision period

In the case of the first PSR, the period has been defined from the beginning of plant operation untilthe date of this first PSR. For subsequent PSR, the period follows the general criteria, that means 10years.

4.3.5.2.2. Revision scope

83

The revision applied to those programmes under development within the NPP as a consequence oftechnological updating needs. The performance of these programmes is based on technologicalinnovation projects, results of research and development programmes related to safety, and therequirements established by the Authorities.

In general, the areas to be considered for the development of specific programmes to be reviewedare those included in the following documents:• CSN Safety Guide 1.10 (see point 4.3.4.4)• Joint Action Plan of Spanish NPPs. Activities related to safety.

4.3.5.2.3. Revision criteria

The criteria to be followed by the NPP in the revision and verification of safety evaluation andimprovement programmes are the following:

• Programme status and compliance level• Programme development at the moment of PSR initiation.• For those finalised programmes:

− Objectives compliance− Activities to be performed and their planing or execution schedule.− Implementation process.− Significant changes in the established conclusions as a consequence of international

state-of-the-art evolution.

• For those programmes under development, the objectives compliance will be considered and ifthe development is according to the foreseen planning.

4.3.5.2.4. Revision procedures

The revision process will take into account the following steps:

• Baseline checking: NPP initial situation, related documentation, references, etc.• Verification of compliance of Authorities requirements, as well as justification of existing

deviations.• Analysis of subject evolution since the start of the programme, as a consequence of results of

new research and development programmes. It will be considered if some programmes activitieshave to be modified due to the above-mentioned evolution.

• Review of activities performed and their planing and implementation status, in order toproduced a programme updating.

4.3.5.2.5. Report contents

For each reviewed programme, a report will be produced, containing a summary of the obtainedresult:

• Foreseen initial programme• Objectives and planing compliance analysis• Modifications performed during the programme• Deviations justifications• Information about the programme status, explaining the actions performed and foreseen• Identification of needed corrective actions, as a consequence of the revision.

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Due to differences in regulatory systems and the interest of evaluating existing practices for oldplant management, the UK approach will be more particularly developed hereafter.

84

Concerning Germany, Ageing management programs exist, but, in the moving German nuclearpolitical context, it is difficult to precise at the moment how ageing problems approach could bereassessed. Current discussions show that plant closure decision will not be justified by ageingmanagement considerations.

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The United Kingdom approach is given with some technical details (not reported elsewhere) thatillustrate the methodology used to maintain a safety level on gas cooled design nuclear powerreactors among the oldest NPP operated in the world.

The 26 reactors of the first generation programme, using natural uranium fuel rods clad in amagnesium alloy, graphite as the moderator, and carbon dioxide as the coolant, known as"Magnox" reactors, were commissioned at 11 sites ; 20 of them are still operating, including the 8oldest reactors at Calder Hall and Chapelcross which started commissioning in 1956.

The 14 reactors of the second generation programme, known as the AGR advanced gas cooledreactors, built at 7 sites, use enriched ceramic uranium dioxide fuel clad in stainless steel ; theirpressure vessels are pre-stressed concrete, and operate at higher coolant pressures and temperatures,giving greater power outputs ; the first of them was commissioned in 1976.

The UK regulatory approach purpose is to ensure that the effects of ageing in nuclear power plantsare fully taken into account at all stages of a plant’s life, thereby giving continued confidence thatsafety is properly addressed.

The nuclear safety licensing regime in the UK is based upon a non prescriptive, goal settingapproach in which licensees of nuclear installations have duties placed upon them through the 35conditions attached to each licence.

The Health & Safety Executive (HSE) is the licensing authority, and the Nuclear InstallationsInspectorate (NII) is that part of the HSE which has the responsibility for granting nuclear sitelicences.

There are two licence conditions which are particularly important in the consideration of ageing ofUK NPPs.

- The first of these enables a specification to be issued which requires any plant on the nuclearlicensed site to be shut down at periodic intervals for examination, inspection, maintenance andtesting, as required by the plant maintenance schedule. When such a specification has beenissued, the licensee must obtain the consent of HSE before the plant can be restarted. This is theprincipal method applied to reactors to ensure that the plant is adequately safe for a furtherperiod of operation.

For most of the reactors the period between shutdowns is 2 years, although the licensees ofsome reactors have successfully argued for a periodicity of 3 years. Since all sites have aminimum of 2 reactors, the periodicity of shutdowns allows consideration of matters whichcould affect the safety of the other reactor at the site to be considered even more frequently.

- Whilst this approach provides considerable confidence in the safety of continued operation, itwas recognised some years ago as the first generation of gas cooled nuclear reactors approached20 years of operational life that a more comprehensive safety review was needed, independentof the pressures of routine biennial reactor periodic shutdown programme. The concept ofperiodic safety reviews (originally referred to as long term safety reviews) was thereforedeveloped, and is the other standard condition of the nuclear site licences important to thesubject of ageing. A programme of reviews has been specified for each reactor site, based on a10 year periodicity.

The principal objectives of the PSR are:

85

i. to review the current safety case for the plant, taking account of plant modifications which havebeen made over the years, the plant’s operational history, and the PSR findings themselves

ii. to undertake a comprehensive review of ageing mechanisms, with an objective of identifyingany potentially long term mechanisms which may limit the life of the plant in the next 10 yearperiod ; actions can be specified to strengthen, where appropriate, the routine inspection,maintenance and testing activities to monitor ageing and degradation effects in the plant

iii. to undertake a comparison of the developments in safety standards since the time of design andconstruction, or within the time period since the previous review, and to identify any reasonablypracticable improvements which should be made to further improve the safety of the plant.

The PSR programme for the Magnox plants started in the early 1980s. Each plant was required tosubmit its PSR for consideration, and the NII published its findings of the initial reviews over theperiod 1987-1995 ; over this period, three of the Magnox NPP, comprising 6 reactors, were closeddown because the safety improvements necessary to enable continued operation was noteconomical.

Early in this programme of safety reviews, a number of generic safety issues were identified whichthe licensees were asked to act upon in advance of completing the reviews of the remaining Magnoxplants. These issues became known as the Generic Issues programme, and the objective was toachieve early resolution of priority issues in order that effective improvements to safety could beintroduced in advance of the completion of a plant’s PSR.

The second 10 year PSRs round for the Magnox reactors have quite been completed with theexception of one, to be submitted to the NII in 2003 ; the findings are mainly related to ageingissues and will be discussed in more detail later.

The UK’s second generation NPPs (AGRs) have also now completed their initial PSRs. Theseevaluations have resulted in programmes of work, including technical studies and major plantmodification work for the older plants.

The NII findings have identified a number of safety issues requiring further attention, likecompletion of identified improvements to plant, further studies to consolidate the long term safetycases for some components in the pressure circuits where they penetrate the pre-stressed concretepressure vessel, and improvements in the confidence in the long term integrity of the graphite cores.

As indicated above, the issue of ageing and the potential for limiting the life of a NPP wasidentified as a generic issue at a very early stage in the PSR process. The NII concluded that somedegradation processes would require more regular reviews than that implied by the 10 yearperiodicity of the PSR. In response to this, the licensees have established generic arrangements toundertake this important work at all installations, and the results of this ongoing programme aretaken into account by the NII when making regulatory decisions such as issuing a Consent for areactor to return to routine operation after its statutory shutdown.

There are several factors which are taken into account in the management of ageing, and primaryexamples can be drawn from the UK’s Magnox reactors. Early in the life of these reactors, oxidationof some steel components was identified as a safety concern. Considerable research into this issuewas undertaken in order to gain a full understanding, and the licensees developed long term safetycases for continued operation.

Inspections of potentially affected components are made at each biennial statutory shutdown andthe results are reported together with the implications for the safety case to the NII prior to thereactor returning to power. This process has been in place from a very early stage of the Magnoxreactors life, and has proved to be a very successful means of controlling ageing effects andcarefully monitoring the actual plant state to confirm that the safety case remains valid.

Another important ageing effect is that of irradiation induced degradation in the steel reactorpressure vessels. Surveillance specimens of the original RPV materials are withdrawn from the

86

reactors at regular intervals to enable various analyses to be undertaken to determine whether thematerial characteristics are still within the predicted limits, and that no previously unidentified, ornew ageing mechanisms are taking place. Although the reactors were not originally designed forin-service inspection, limited inspections are undertaken at each statutory shutdown to confirm thatno unexpected conditions have developed. These inspections are targeted at areas of the RPV wheredefect tolerance is least good, such as the outlet duct nozzles. This has required the development, bythe licensees, of special non-destructive testing equipment capable of carrying out the requiredinspections in remote parts of the RPV.

The safety case for the RPV is supported by detailed structural analyses, which are aimed atproviding assurance that the vessels can tolerate very large defects without failing, setting operatingconditions so that even if a defect were to develop and penetrate the RPV wall it would not lead tofast fracture, and would be detected by the carbon dioxide leak detection system before it couldgrow to a critical size.

These analyses are regularly updated to incorporate the latest materials surveillance data, and arealso used to determine the temperature limits which must be maintained in order to keep the RPVmaterial in a fully ductile state during routine operation. Following the closure of the TrawsfynyddNPP a few years ago, material property prediction methods have been validated against samplestaken from the reactor pressure vessels to provide additional assurance.

The graphite used as the moderator is subject to various ageing mechanisms, including irradiationinduced effects and other effects such as erosion. As in the case of the RPV, the integrity of thegraphite core is kept under regular review and is reported upon prior to consent being grantedfollowing the statutory shutdown. Graphite samples are routinely removed from the reactor cores,and are analysed to determine whether the properties are still within the predicted limits. This isfurther supported by remote inspections of selected channels of the core to determine whether therehas been any significant graphite distortion or movement which may inhibit the entry of the safetyshutdown control rods. Finally, the results of control rod insertion tests are also examined toconfirm that there is no impediment to control rod entry.

The above discussion has been centered primarily on ageing factors specific to the Magnox RPVssteel and the graphite ageing factors generic to gas cooled reactors. There are many other items ofplant important to safety to which attention is also given as part of the ongoing maintenance,inspection and test programme, and which have similarities with other reactor types. As part ofNII’s requirements placed upon the reactor licensees from the periodic safety reviews, moreinformation from maintenance activities is now being gathered in order to provide data for trendanalysis and indicators for ageing effects in such plant.

Finally, it is worth noting that the industry maintains a research capability to support its indigenousgas cooled reactor designs. This is achieved by the licensees having their own researcharrangements, either within their organisation, or in collaboration with other bodies, includinguniversities with the required expertise. In addition, the HSE has the responsibility for ensuring thata balanced nuclear safety research programme is in place and that adequate research projects arecommissioned. This is administered by HSE’s Nuclear Safety Directorate, which also has the powerto require research to be commissioned which is then charged to the Nuclear industry licenseesunder a levy system.

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The main European Union countries which have nuclear power plants have a quite commonapproach of regulatory requirements related to NPP lifetime management.

In particular, there are no limited time operating authorisation, and safety is the responsibility of theutility, under continuous surveillance by the regulatory authority.

87

However, according to the different countries, the authorisation given by the safety authority to theplant operator is not associated to the same formal process. It depends on the safety authority, itsorganisation and its relations with the nuclear power plant operators.

Formal Ageing management evaluation processes exist in some countries, for quite short periods(i.e one year in Spain, two in the UK) ; in others, it appears through a requirement of ability forsafety demonstration at any moment (in France and Belgium).

In practice, three ways of safety ageing management are implemented.

− The first, clearly expressed, is the periodic safety review (PSR) approach, widely accepted inthe international community, even if it is not always required by the regulation ; a ten yearsperiodicity is also a common practice, at the moment.

This is the principal method applied to reactors to ensure that the plant is adequately safe for afurther period of operation. But it appears a need to strengthen this periodic approach by acontinuous activity of ageing surveillance and management, taking into account safety andindustrial anticipation needs.

− The second is often covered through the implementation of a Life-time ManagementProgramme, often less accurate in its safety concerns.

However, it could be the framework for an answer to the complementary need of an effectiveand efficient approach, periodically reporting ageing safety issues developments andimprovements, able to prove that safety margins are maintained.

− The third one is sometimes constituted by generic evaluations of specific important issues,which may be covered by adapted procedures, with their own approach and program: this mayconcern cases like vessel failure assessment, stress corrosion cracking of 600 alloy, and thermalageing of austenitic-ferritic cast parts.

Comments on these three aspects are given in the following paragraphs.

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Periodic Safety Reviews are applied to numerous plants in different countries, and provide acomprehensive methodology for reviewing the safety of the plant, including ageing effects andregulatory evolutions, technical progress and technological obsolescence.

Although the expression of objectives and practical evaluations may slightly differ, there is a globalconsensus on PSR objectives and content. The following aspects are considered in each country:

− the safety codes and standards evolutions, including international recommendations, and theidentification of improvements to be made,

− the review of potential ageing mechanisms, with the objective of identifying effects likely tolimit plant life during the next ten years period, and defining necessary actions on use,surveillance, maintenance or mitigation measures,

− the operational history of SSC, the in-service inspection results, and the lessons learn fromreturn of experience on similar SSC. There is a consensus on the credit to be given to theexisting programs and knowledge of component history for the definition of the subsequentsurveillance programs or license renewals.

The following aspects are more specific to some countries:

− formal updating of Lifetime Management Programmes in Spain,

− the results of Probabilistic Safety Analyses, which are more explicitly considered in Spain,

− more systematic reference to the regulatory evolutions in the project origin country in Belgiumand Spain,

88

− the evaluation is generally based on the initial licensing basis in Spain and Belgium, exceptwhere justified by return of experience, where the French approach is more oriented to ensureconsistency of approaches used for the various plants, thus referring to the last status of rules.

In practice, nevertheless, upgrades are generally applied where reasonably achievable, on a case bycase basis.

The technical evaluation of the different ageing phenomena may include specificities in the variouscountries, depending on the expression of the regulatory objectives, for example:

− definition of the integrity objective concept: acceptability of crack-arrest phenomenonconsideration for defect acceptance, definition of fatigue initiation objective [73], etc.,

− prescribed safety margins (prescriptions are given in this field in the French November 10, 1999order), which are implicitly related to an acceptable level of risk,

− differences on consensus on technical aspects such as warm-presstressing effect or acceptanceof conditions leading to significant loading in the brittle-ductile transition, etc.

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In every country, there are ageing or "service-life" management programs. Their status isnevertheless different:

− in France and Belgium, it is under the responsibility of the Utility, which must operate safelytaking also into account the economic aspects in order to reduce the operation costs. From alicensing point of view, the utility shall be able to demonstrate at any time the safety of itsplants,

− in Spain, the license requires that each year plants owners prepare and submit to the RegulatoryAuthority an annual report including a Lifetime Management Programme, its status as well asactivities performed within this Programme. As a result, plant operators generally followUNESA methodology,

− in UK, a formal consent is needed to restart the plant after shutdown due to maintenance orrefuelling, providing the opportunity to ensure, with a 2 to 3 years interval, that the plant isadequately safe for a further period of operation. Generic issues are considered in particular.

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Due to the standardisation of plant series in France, and the potential importance of generic ageingphenomena, since 80% of electricity is generated by NPPs, there is a strong economical interest forthe Utility, and a strong strategic interest for the Safety Authority to anticipate these potentialproblems.

Generic studies have consequently been conducted by the Utility to support the decisions taken foreach individual plant, with a specific management and periodic discussions with the safety authorityand its technical support.

Examples of such generic studies are:

− the re-evaluation of the resistance to fast fracture of the RPV,− the susceptibility to stress corrosion cracking of all 600 alloys parts,− the thermal ageing of austenitic-ferritic castings.

These studies consider the R&D progresses, the results of in-situ examinations, the detailed analysisof construction conditions, the service conditions, and lead to optimised in-service inspection andmaintenance proposals, taking into account the possible anticipations through the examination of

89

"precursors" deduced from the conclusions of these studies, even where such precursors may not beimportant for plant safety by themselves.

90

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In Belgium, many aspects of plant life management have been incorporated in the every daymanagement of the plants since the beginning of their life. These aspects include: the design, thequality assurance and control, in-service inspection, monitoring, testing, preventive and predictivemaintenance, re-qualification, replacement, periodic safety reassessment, etc.

A specific plant life management project, named "Continuous Operation of Belgian NPP’s" also hasrecently started. Its main objective is to centralise all safety and economic aspects of plant lifemanagement. The two main outputs of the project will be:

• the determination, for each NPP unit, of the most probable cost required to maintain safely theunit in operation;

• the identification of the actions to be taken (e.g. predictive maintenance, monitoring) to achievethat at the minimum cost.

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In order to focus on the most significant Systems, Structures, and Components, the components areclassified into four categories:

− active safety-related components,

− passive safety-related components;

− non safety-related components but components important for the availability of the plant;

− and non safety-related components and not important components for the availability of theplant.

For the active safety-related components (e.g. motors, pumps), maintenance programs wereestablished. These programs define the actions to perform in order to guarantee the integrity and theavailability of these equipments during the exploitation of the NPP. Thus, a low priority was givento these components and they are excluded from the scope of the current ageing managementprogram2.

A very low priority was also given to the last category (i.e. non safety-related components and notimportant for the availability of the plant).

Hence, the current ageing management program focuses on the second and the third categories. In afirst phase, Equipment Ageing Summaries, EAS, were established for the passive safety-relatedcomponents. In a second phase these EAS will be extended to the major components important forthe plant availability in order to address also economical concerns.

Each Equipment Ageing Summary concerns an ageing phenomenon applied to a System, aStructure, or a Component. These EAS provide a status on the activities presently being undertaken

2 : An EAS was however established for IE electrical and I&C equipment describing the qualification

programs

91

to follow up ageing phenomena of components in the Belgian Nuclear Power Plants. The typicalEAS table of content is given in Table 5.1.

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1. Equipment2. Problem description3. Problem category4. Problem significance5. Significant parameters6. Sketch or drawing7. Risks 7.1. Safety risks 7.2. Economical risks8. Surveillance 8.1. Type 8.2. Frequency9. Acceptance criteria10. Remedies 10.1. Repairs 10.2. Replacement unchanged 10.3. Replacement with redesign 10.4. Other remedies11. Cognisant engineer12. Participation to international working groups13. Main references14. Situation in Belgian units15. Follow-up by Belgian Safety Authority16. Project references

After identification of ageing phenomenon-SSC pairs, a ranking will be performed in order to focuson the most important pairs.

The method that will be proposed in Belgium to rank the pairs is to consult experts from the utilityand from the engineering company. These experts will rate prioritisation criteria (See examplesgiven below). The results for each category will be combined taking into account the relativeimportance of the criteria and a ranking will be done.

Examples of criteria are given below.

− High potential (probability) that ageing degradation causes component failure;

− High susceptibility of the component to ageing;

− High replacement costs in case of failure (i.e. component cost, plant shutdown);

− Low replacement costs but high impact for the plant operation cost in case of failure;

− High impact for the risk level of the plant in case of failure (e.g. use of PSA studies);

− Probability to detect the ageing process before plant integrity is significantly degraded;

− Availability of data from direct observations on the plant (or from elsewhere if not available) topredict ageing;

− Level of confidence in future predictions;

− Feasibility and effectiveness of current practices in the industry in mitigating ageing (repair,replace);

− The component could become obsolete with an inadequate availability of spares;

92

− The component has required significant maintenance or / and has been replaced during itslifecycle;

− Operation or environmental conditions are different than those considered in the design.

From the ranking, the most of important ageing phenomenon-SSC pairs will be identified andincluded in the ageing management program and new EAS will be established.

For the identification of the important pairs, the following criteria will be used:

− all safety-related components will be included in the plant life management;

− non safety-related components will be included in the plant life management if their rank isabove a determined threshold.

The ranking will be performed for each Belgian Nuclear Power Plant to take into account the plantsspecificities. A first general tentative ranking of existing EAS (i.e. for passive safety-relatedcomponents) was done and is given in Appendix 2.

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Based on experience feedback from the Belgian and foreign nuclear power plants (e.g. technicalmeetings, articles, conferences), ageing mechanisms for components whose failure or malfunctionmay prove hazardous for the safety were identified. For each ageing phenomenon-SSC pair, an EASwas established. The list of all existing EAS is given in Appendix 3.

Up to now, the content of the EAS addresses only the safety aspects; i.e. only the componentsimportant to safety are covered and the risk description is limited to safety risk. It is now intendedto create new EAS and to complete them in order to address also economical concerns. This willproduce new outputs, mainly the determination for each NPP of the most probable cost required inorder to keep economical efficiency as well as safe operation.

To identify all possible ageing mechanisms, Tractebel has conducted bibliography search and hasestablished a list of main reference papers and reports. The main information of these articles iscollected in a database: the "Plant Life Management Documentation". Also technical meetings willbe organised with experts from both the Utilities (Electrabel) and the engineering company(Tractebel) in order to complete the list of ageing phenomenon-SSC pair.

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Examples of rules for ageing prediction and of acceptance criteria are given in Appendix 3 of thisdocument.

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Surveillance includes in-service inspection, monitoring, testing, etc.

All classified pressure retaining components are inspected according to the requirements of theASME Code Section XI. Moreover, complementary or voluntary inspections are decided by theoperator on classified and non classified components according to variable aspects affecting theavailability and the conventional security of the plant or depending on world feedback experience.

Monitoring is also used in several instances. For example, the monitoring in type and frequency oflow cycle thermal and pressure transients in order to control components subject to fatigue andeventually in order to justify the extension of their fatigue life.

More examples of surveillance and periodic testing practices in Belgium are given in Appendix 3.

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93

Examples of methods used to mitigate ageing effects in Belgium are described in Appendix 3 forselected Equipment Ageing Summaries.

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Examples of remedies applied in Belgium are given in Appendix 3.

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The inventory of pressurised water reactor power plants (PWR) represents technical and financialcapital of strategic importance, for the Owner and for France. It should be borne in mind, inparticular, that 58 PWR units provide nearly 80% of national electricity production.

The first main specificity of the French nuclear power plants is their standardisation. 34 CP0-CPY900 MWe, 20 P4-P’4 1300 MWe, and 4 N4 1450 MWe plants have been built. This quitestandardised design and construction practice leads to risks of generic ageing problems, but also topossibility of in-depth generic studies on the significant ageing phenomena, and to define an in-service surveillance based on the evaluation of "precursors" defined following these studies.

The second specificity results from the high percentage of French electric production being issuedfrom nuclear power plants. This leads to the necessity of a power network follow, and consequentlyto specific potential fatigue ageing problems, and to a particular need for a continuous transientmonitoring.

The third specificity is the existence of only one major Utility permitting an harmonised lifemanagement at the set of plants inventory level.

Chapter 4.2 has presented the consequences of these specificities on the safety organisation. Thisapproach may be summarised as follows: a continuous technical dialogue between Utility andSafety Authority, the ability to make a safety demonstration at any moment (following for examplereturn of experience), a periodic safety review in phase with the decennial inspection visits, with apossible reduction of this interval to a 5 years periodic evaluation at end of life..

The amount of electricity generated by nuclear power gives a strategic importance to a goodmanagement of the life duration of the plants, this consideration having a strong influence on futureplant construction needs.

For the past 20 years, the choice of the nuclear programme has allowed a large margin ofcompetitiveness for production, to ensure energy independence and to reduce CO2 emissions. Forthis reason, for the next 20 years, the service life of nuclear power stations will be a deciding factorsince design studies made it possible to contemplate a service life of 40 years. It must be stressed,however, that this service life is not regulatory in nature and does not constitute a guarantee. TheUtility must therefore do all what is necessary from the safety and public acceptance point of viewsto justify and attain this service life, and if possible, to extend it in order to make the best use of theinvestments already made.

The service life of a nuclear power plant can be affected by three factors:

− Normal wear and tear of its components and systems – sometimes called ageing – whichdepends, in particular, on their age, operating conditions and the maintenance operations theyreceive,

− The safety level, which must always conform to the safety level of reference applying to theplant under consideration, and which is likely to change with new regulations,

94

− Competitiveness, which must remain satisfactory, compared with that of other means ofproduction.

In practice, these three factors are closely linked. Ageing can degrade the performance of the powerstation and its safety level, and attempts to correct this problem create preventive and correctivemaintenance costs that must be controlled. It is also necessary to pay attention to the industrialenvironment and to the strategies chosen insofar as maintaining knowledge, skills andmanufacturing capacity is necessary for long term operation.

Finally, the service life of the inventory in operation has direct repercussions on the plantreplacement strategy.

The topics linked to the nuclear plant service life can therefore be summarised as follows:− understanding and anticipating ageing issues,− maintaining the required level of safety and performances,− controlling costs in order to maintain competitiveness,− ensuring the better use of investments already made, and manage appropriately future

investments linked to renewal of the existing production inventory,− maintaining the public’s trust.

These stakes prompted the owner to deal with ageing issues very early on, and to start, in 1985, avast study and research programme, which was named "Service Life" [108]. The "Service Life"programme had the following objectives:

− Identifying the different degradation phenomena likely to affect material and components, usingnational and international return of experience,

− Understanding the different ways in which these phenomena appear and the ways in which theyspread (their kinetics),

− Identifying the components that may be subject to these phenomena, the replacement of whichis impossible and for which all steps necessary must be taken to provide proof of theirsatisfactory behaviour over the course of time,

− defining a maintenance policy for other materials, making it impossible to level out thedegradation phenomena, that can be prevented by monitoring, repair or replacement.

This programme which combines all the skills of the owner and its main partners (manufacturers inparticular), brings into play several fields of expertise in order to guarantee coherence andexhaustiveness of thought. It is, in addition, subject to extensive international exchanges. Studiesundertaken have had to take into account design and manufacturing data, operating conditions,maintenance strategies, to rely on R&D actions, and, finally, to create a suitable industrial policy.

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For each zone evaluated for the preparation of in-service surveillance programs, an evaluation ofthe gravity of the consequences of a potential failure was made, based on the "engineeringjudgement" of experts participating in the working groups. These consequences were classified inthree classes: low, mean and high. The combination of the global number representing theprobability of occurrence of a damage as determined in 5.2.3, and the class of its consequencespermitted to identify inspection requirement classes, according to which a global appreciation of thecurrent examination practices (as in 1976) was conducted, identifying possible needs forimprovement.

Four level of inspection needs were defined:− no particular examination needed,− limited examination accepted,

95

− mean examination level,− satisfying examinations required.

The "Service Life" program was more focussed on replaceability of components, identifying:

− Replaceable components, entirely or partly, sometimes at the expense of large operations. Thisis the case, in particular, for steam generators (SG of seven units have already been replaced),closure heads (30 have been changed), parts of instrumentation and control, generators, etc.

− Two components declared irreplaceable, the reactor vessel and containment, for which allpossible measures must be taken to prove their satisfactory behaviour over the years.

Components considered on the basis of the routine and/or exceptional preventive maintenance arelisted in Appendix 2 with their prioritisation. Only the vessel and containment are considered asirreplaceable.

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At the very beginning of the seventies, for the preparation of the in-service surveillance rules,various potential damages were identified on return of experience and engineering judgement bases.Factors related to mechanical or thermal loading (geometrical discontinuities, external loads, forces,vibrations, thermal gradients and residual stresses), to fabrication and examination (potentialdefects, metallurgical properties, manufacturing and examination difficulties) and to theenvironment (corrosion due to a low fluid speed, at a steam-water interface, due to local boiling,erosion, abrasion, matting, tearing, impact and shock possibilities, irradiation) were examined.

For the particular case of the Main Primary System, 93 zones with 296 sub-zones were examined, anote being given for each sub-zone to each of these potential damages: 1 for a very small to smallrisk, 2 for a small to mean risk, and 3 for a mean to high risk.

A global (arbitrary) appreciation of the damage risk of each zone was based on the sum of the abovenumbers. This permitted the identification of the "weak" zones of the system for which a particularevaluation of the available non-destructive examination methods was needed.

After an average of 18 years of operation for the 34 units of 900 MWe, and 12 years for the 20 unitsof 1300 MWe, the French units may now be considered young and mature. A more in-depthevaluation of the various ageing mechanisms was made more particularly for components difficultor impossible to replace in the context of the "Service Life" study presented in 5.2.1.Accompanying sheets are given in Appendix 3 for the degradation mechanisms likely to affect:

− the reactor pressure vessel,− the reactor containment,− cast parts used for pipes, bends and primary pump casings,− instrumentation and control,− electric cables.

Large study programmes have been devoted, since 1985, to the understanding of the mechanismsand kinetics of degradation: embrittlement by irradiation, erosion, stress and fatigue corrosion, etc.They allow, first of all, maintenance and operating policies to be optimised.

To confirm the hypotheses resulting from this work and to validate the non-destructive test resultsby additional inspections, expert appraisal programmes on materials taken from operation havebeen designed: they contribute to a better understanding of the updated phenomena. In this way, alarge programme has begun on the Chooz A plant (300 MWe), the first PWR built in France, shutdown in 1991, after 24 years of operation. Appraisals were carried out, in particular, on the reactorvessel and the internal structures, in order to better determine the effects of ageing under irradiation.

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96

The national and international return of experience following service problems or resulting from theabove approach is closely followed, and potential consequences on not (yet) affected plants isevaluated through generic ageing studies.

A close analysis of the specificity of the part, material, loading environment, etc., permits toidentify "precursors", to predict their ageing, and to verify through a close examination thesepredictions, permitting the definition of a global strategy at the level of the set of French NPPs.

Examples of such approaches may be found for the fast fracture evaluation of the Reactor PressureVessels (RPVs), the Stress Corrosion Cracking in Inconel parts, or the thermal ageing of austenitic-ferritic cast parts.

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The evaluation of the existing practices together with the damage probability/consequence balanceevaluated at the beginning of the seventies did lead to examination practices discussed with theSafety Authority. After more than ten years of experience, a working group was established tostandardise the resulting practice. This has led to the 1990 edition of the RSE-M [60] coveringmechanical components, which is periodically updated. Other RSE standards are being prepared onother equipment.

The acceptability of defects discovered is subjected in particular to detailed evaluation proceduresdescribed in this code, which take into account their potential evolution during the subsequentperiod separating two examinations. Acceptance criteria remain deterministic, but are based onreliability considerations, taking into account actual properties distributions, and include prescribedsafety margins.

Surveillance include continuous monitoring and partial or complete evaluations during shut-downperiods.

In view of the present state of knowledge, it appears in the context of the "Service Life" programmethat in return of operating conditions, suitable monitoring and maintenance of components, thenuclear units should reach the desired objectives with regard to service life.

97

Regarding, in particular, the two irreplaceable components:

− The reactor vessel, owing to low content of residuals (copper, nickel, phosphorus) andoptimised management of fluence, thus limiting embrittlement of the materials underirradiation, should reach 40 years without special operation. However, the studies have clearlyshown the need for optimised irradiation monitoring and additional actions listed in Appendix 3in order to validate and reinforce this service life potentiality.

− Monitoring by auscultation and reactor containment pressurisation routine tests show that someof them can have strength and/or leaks resistance problems. These problems result fromdeformations by shrinkage-creep, greater than that taken into account in the design.

Moreover, the behaviour of the containment can differ considerably from one site to another,depending on the aggregates used during construction. An individual estimated follow-up istherefore necessary (See appendix 3).

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The methods used to mitigate ageing effects in France are given in the accompanying sheets (SeeAppendix 3).

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The industrial strategy is primarily function of the classification of the component according to itsreplaceability. The maintenance policy adopted for replaceable components may be, firstly, tomonitor, repair a component before deciding to replace it when the performance-safety-costsappraisal requires this. This is, in particular, the policy adopted for steam generators.

A reinforcement of the anticipation approach in exceptional maintenance has been decided on thereplacement register of some components: the approach consists in demonstrating their operationalreplaceability ranging from the feasibility study to making available skilled servicing facilities.

Beyond continuous checks, during operation and shut-down periods, it was agreed with the safetyauthority, within the context presented in 4.2, that every 10 years a re-examination of conformity ofthe units per plant series would be carried out. It is also an opportunity to reassess the referenceframework of the safety requirements itself according to the knowledge acquired, feedback ofexperience, technical progress and to carry out a number of additional checks. The developments innational and international regulations (basic safety regulation on earthquakes, restriction of waste innormal and accident conditions, etc.) are considered for the development of the safety requirementreference framework.

The 900 MWe units have thus been subject to a reassessment leading to a new safety referenceframework which should remain stable for 10 years. It leads to technical modifications approved bythe safety authority: hazards (fire, cold, earthquake) protection devices, automatic response tooperating incident systems, or procedure improvements. These modifications are carried outconsistently during the second decennial outage programs (VD2) of the 900 MWe plants, the first ofwhich has taken place in 1998 at Tricastin 1. A detailed check of installations, maintenanceoperations and a set of safety improvements are conducted. Moreover, the VD2 of the 900 MWeunits are the opportunity to carry out an additional action programme in order to check, by controls,hypotheses made on the absence of degradations in the non-controlled zones by way of thepreventive maintenance programmes.

A new safety reassessment will be programmed some years before the third decennials (VD3) thatthe older 900 MWe units will reach towards 2007. The third decennial will be very important forthe plant service life and, consequently, its preparation will be essential: it will be necessary to havea file demonstrating very explicitly the control of safety and reliability of the main components andsystems.

98

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As already stressed, the ageing of components and systems making up a nuclear power plant canhave repercussions on the performance of the installation and on its level of safety (and therefore itsservice life) and must be overcome by adapted maintenance actions.

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− Monitoring during operation and unit shut-downs.

− Repair of an existing piece of material, each time that it is possible to intervene at an acceptablecost and dosimetry on this material in order to restore its operating characteristics. These actionsare considered as routine.

− Replacement of a material for which the projected wear, therefore in the future, no longerallows its operating characteristics to be restored. This is essentially "exceptional maintenance".

The choice between these types of actions is not only technical. It is also economical, because theowner must, in a sufficiently prospective approach, be able to optimise the material maintenance byanticipating the time when repair will become too expensive compared with the cost of areplacement.

The development of new monitoring techniques permits a conditional rather than systematicmaintenance, the use of stricter analysis methods in order to design programmes (Optimisation ofmaintenance by reliability data) and the creation of anticipatory replacement strategies, allowingcosts to be controlled, while efficiently maintaining the safety and reliability of the units.

The second competition factor lies in the reduction of length and the dosimetry of the actionscarried out during unit outages performed every 12 and 18 months. These outages, intended topartially replace the fuel and to carry out routine or exceptional (replacement of steam generators orclosure head of a unit, for example) maintenance operations require perfect planning andorganisation of actions and participants. This is, of course, even more important during decennialoutage programs, where the number of operations to be carried out and of participants is far greater.

Significant progress has already been demonstrated with regard to unit outages conditions. In thisway, for example, the replacement period for steam generators is set at less than 35 days, with awell controlled dosimetry of 1 h..sv.

Among the other actions carried out to increase the availability of nuclear power plants, one shouldalso mention those undertaken for some years to increase the length of reactor operating cycles(length between two outages to partially reload fuel), This cycle length which was planned at 12months in the design is today 18 months for most 1300 MWe units: This cycle increase could begradually extended to all 1300 MWe units and to CP0 (Fessenheim, Bugey), after the agreement ofthe safety authorities. The availability of the inventory which was greater than 82% in 1997, shouldthereby earn some points and reach 85%, the objective set for the year 2000.

Another essential factor of nuclear ageing management is industrial control. Whether this concernsnational or regional suppliers which are contributing to the plants, specialised technical skills,necessary for engineering the park in operation, or the manufacturing of the most sensitive materialswhich require high-tech, unique industrial tools, the industrial and supply policy must be in linewith the duration and for that must employ a partnership policy with suppliers without excludingjudicious competition between suppliers.

Through the results achieved over the past years in the context of the service life studies, the owneris today able to better anticipate and overcome ageing phenomena, and, consequently, to maintainits installations in a condition allowing a safe service life of at least 40 years.

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99

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In Spain, NPP owners are increasingly concerned with optimising Plant Life Management. Inresponse, they are setting up Lifetime Management Programmes.

Strategic, economic and safety concerns and the close link between life management work and theimproved maintenance practices that are so important today, will increase and globalise theseprogrammes for monitoring and conservation or mitigation of ageing.

These programmes are all based on knowledge of the precise condition of all components andpopulations with the greatest effect on the economics and safety of the plant, and trends in changesin their condition.

The technical support for these programmes is:

− Methodologies and knowledge required identifying degradation mechanisms as a function of thecharacteristics of the components or populations, and servicing conditions.

− Techniques for determining condition and trends over time.− Analysis of the efficiency of maintenance practices based on the above knowledge, techniques

and methodologies.− Improvement of maintenance practices for adequate mitigation and monitoring of ageing.− Techniques and tools for collecting and ordering data about ageing and for condition assessment− Exchange and feedback of domestic and foreign experience in lifetime management / ageing

programmes, considering international practices and recommendations.

The prediction of potential ageing and evaluation of the degree to which this affects the differentcomponents, and especially the monitoring of change in their condition and/or prediction of thechange, as well as the definition of corrective or monitoring measures, require specialisedengineering support in these fields.

Figure 5.1 illustrates the structure and content of the Unesa methodology for NPP LifetimeManagement. The following sections describe its structure and content, with special emphasis onengineering tasks.

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This paragraph provides a summary of the UNESA standard method for selecting and prioritisingplant components important to lifetime management of LWRs. Additional details on screeningmethodology and criteria definition are given in Appendix 2.

The first requirement for adequate plant life management is to avoid dispersion and waste of theLCM Programme resources. These resources are always limited, and should not cover the wholepopulation, indiscriminately. Prioritisation is necessary and needs to be slightly adjustedperiodically, to adapt to the margin of uncertainty of all predictions. This prioritisation uses aweighted criteria methodology.

Components having the greatest sensitivity to ageing, and operating and maintenance (O&M) costswill dictate the feasibility of Lifetime Management Programme for the plant and influence utilitystrategic planning. These components are termed the "important components” and will be the objectof further investigations and research to identify the parameters which affect their life cycle.

100

The strategy of each plant affects the methodology through adjustments in the plant-uniqueweighting of each of the criteria.

The selection and prioritisation methodology (Figure 5.2) is established in three steps:

− In the first step the systems are selected according to safety, availability and cost criteria.

− In the second step, the selection of components from the systems is performed based on a widerange of criteria such as:

. Component is safety-related or required for safe shut-down,

. Component failure has a significant impact on safety level,

. Important component in licensing process,

. Operating and environmental conditions are more aggressive than considered in design,

. Significant effort in maintenance is required,

. Component maintenance is not effective for ageing control and mitigation,

. High cost or long period needed for component replacement.

− In the third step, the components are ordered following several criteria such as: serviceconditions and history, regulation factors, reliability considerations, programme effectiveness,etc. Different weights are assigned to the above factors and criteria.

The establishment of grouping criteria allows inclusion in the same class of components withsimilar surveillance parameters and residual life evaluation methods. Therefore, the main reasons togroup them is to make easier the ageing surveillance and management, based on similar parametersand techniques.

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Lists of important component for PWR and BWR plants, according to this methodology, are givenin appendix 2.

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This task (Figure 5.3) consists of clearly differentiated stages. The LCM Programme begins with aninitial condition evaluation, which serves as the basis for establishing the main corrective andmonitoring actions, and for preparing the first cost/benefit analyses for Life Management. The LCMProgramme continues to progress with periodic re-evaluation of condition to confirm the correctivemeasures are the right ones and to adopt new measures, if necessary, as a result of the monitoringestablished.

The initial evaluation begins with a determination of potential degradation mechanisms and of thelevel of harshness of these on the selected components. Determining the importance of degradationmechanisms requires a study of the characteristics of the components (design, materials,manufacture, process and service conditions). This analysis is complemented by a rigorous study ofthe history of the operation and maintenance, and the results of diagnosis and monitoring, to detectincidents that might have affected the condition of the plant, or for evidence of degradations.Uncertainty about the severity of some of these ageing effects may require extra inspections or tests,to provide more precise data. The result of this analysis is an Evaluation Report for each componentor group of similar components.

Condition evaluation requires collection and ordering of the documentation and records ofmanufacturer, operation and maintenance that contain information needed for the analysis. Thiscollection requires application of procedures that establish the data and records, with the periodicityof their acquisition clearly identified for successive re-evaluations, and the screening requirementsfor easier collection and analysis.

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The rules, parameters and tools for SSC condition evaluation and ageing prediction, and the criteriato fix on the limits for alarm or safe operation are given in Appendix 3 of this document.

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The practices in Spain covering the inspection and periodic testing required by TechnicalSpecifications, required by applicable codes and standard, and Plant specific surveillance programsare shown in Appendix 3 here in included.

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In addition to the improvements in operation and service conditions, a substantial part of the causesand effects of ageing mechanisms have to be mitigated by maintenance work. The nature of theselong-term ageing mechanisms has meant that, in certain cases, current maintenance practices do notprevent them. This requires these practices to be evaluated and modified where necessary toimprove their efficiency in conservation and the mitigation of degradation.

The engineering activities (Figure 5.4) followed in the evaluation process are:

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The tasks described above produce the component-degradation mechanism pairs that it isconsidered necessary to evaluate.

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A component degradation sheet (Figure 5.5) is completed for each component selected.

The data to be filled out on the CDDSs are: component description; functions; design parameters;operating experience; degradation mechanisms; and the part of the component affected by ageing.

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Description of the Structure, Component or Component Group References

Required Functions References

Pertinent Design Data/Information References

Operating and Maintenance Experience References

Applicable Stressors Affected Subcomponents/Parts Assumptions/Justifications/Remarks

Degradation mechanisms Affected Subcomponents/Parts Assumptions/Justifications/Remarks

Performance Criteria

Supplementary Remarks

Prepared by: Checked by: Approved by:

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For each of the programmes, practices and procedures that affect each component/degradationmechanism pair, a data sheet (Figure 5.6) is prepared, showing the following information about thepractice: limitations on performing it, time when corrective action is taken, the data necessary,action to be taken to mitigate, detect and monitor the degradation and finally comments andexperience resulting from the practice application.

The purpose of this task is to take and inventory of all practices current at the Plant and to discoverdetails of the application, to exploit them and improve their efficiency for life extension.

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Programme/Procedure/Practice Document No.

Purpose of the Programme/Procedure/Practice

Programme/Procedure/Practice is Documented and in Current Use? Yes No

Structures/Components/Component Groups Covered Limiting Conditions Affecting Scope or Use

Criteria for Corrective Action

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106

Each Component Data Sheet is attached to all the Maintenance Practice Data Sheets that affect thecomponent. With the information from both sources, the Maintenance Evaluation Checklist (Figure5.7) is completed.

0DLQWHQDQFH�(YDOXDWLRQ�&KHFNOLVWPlant Name

Structure, Component or Component Group Plant Identification No.

YESYES WITHCOMMENT NO

NOTAPPLICABLE

SUPPORTINGREMARKS

1. The significant degradation mechanismsidentified on the component degradation datasheet will be detected as evident on themaintenance practices data sheet.

2. The features and/or action steps (content andfrequency of the plant programmes will besuccessful for detecting ageing and monitoringperformance

3. The subcomponents/parts which could affect therequired function(s) are addressed

4. The thresholds/criteria for corrective action willresult in timely mitigation of age-relateddegradation and restoration of degradedperformance

5. Data recording requirements are acceptable tosupport trending, etc

6. Plant programmes and implementing proceduresare documented and in current use

7. Maintenance effectiveness is demonstrated

8. Issue Resolution Forms have been prepared


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The evaluation shows the possible deficiencies in control of ageing of the maintenance of eachcomponent. When necessary, improvements to maintenance are proposed, documenting the detailsof the improvements using the tool developed for that purpose, the Maintenance EvaluationProposed Improvement (Figure 5.8).

Samples of the most common practices in Spain to mitigate the degradation of the affectedcomponents and structures are given in appendix 3.

Changes in service condition, control of fluid chemistry or environmental conditions. Improvementor recovery of material characteristics, and modification in operating modes are the most extendedactions.

,668(�5(62/87,21�5(48(67

PLANT NAME

107

AFFECTED STRUCTURE, COMPONENT OR COMPONENT GROUP IDENTIFICATION NO.

AFFECTED PROGRAMME/PROCEDURE/PRACTICE DOCUMENT NO.

ISSUE DESCRIPTION

RECOMMENDATION(S)

RESOLUTION

REFERENCES


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The tasks described above provide information about ageing and trends, and the degree ofuncertainty in their evaluation, and also a determination of the efficiency of maintenance practicesand their shortcomings. On the basis of this, it is possible to decide on the life extension measuresto be applied. These measures fall into the following categories:

• Repairs, replacements or modifications and most efficient programming, of the componentsmost severely effected and/or for which the improvement in availability or performancejustifies the investment. It is important to remember that residual Life is only considered assuch if it is safe (reliable) and economically viable

• Modifications to operating procedures and/or in service conditions to make them less harsh

• Improvements to Maintenance Practices, to achieve full efficiency, for safe and economicallyviable life extension

• Implementation of additional monitoring with some of the following criteria:

− Improve precision of condition evaluation and trends, for those component /degradation mechanism pairs for which forecasting is more uncertain

− Allow for continuous condition monitoring, or at least to reduce the effortrequired for collection and analysis of the information required during re-evaluation

− This improves the flexibility and solvency of life management decisions

108

− The type of monitoring and the parameter to represent ageing should be selected withrealistic criteria of accessibility and efficiency

The maintenance evaluation methodologies to determine the weakness of the current maintenancepractices, the suitability of their frequency and acceptance criteria to control the evolution of thedegradation effects are shown in Appendix 3 of this document.

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The comparison of the practices of the three countries involved in this project, as described byTractebel for Belgium, EDF for France, and Unesa for Spain, is covered in the followingparagraphs.

Organisational aspects are discussed in 5.4.2 and compared in greater details in Section 6, using theformat of international IAEA recommendations.

As far as the technological aspects are concerned, the practices may be compared at three levels :

• the approach for prioritisation of systems, structures and components,• the practical organisation for identification and selection of ageing mechanisms,• the management of ageing for given components and damage mechanisms.

These aspects are covered in 5.4.3, 5.4.4 and 5.4.5 respectively.

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The first main difference between countries lies in the global ageing management of plants. Twoapproaches are identified:

• The first one corresponds to the Spanish practice, where a dedicated organisation is devoted toageing management, with annual reports identified in the licensing process and consequently aprecisely defined structure,

• The second one corresponds to French and Belgium practices, where the various aspects ofPlant Life Management have been incorporated in the every day management of the plants, witha periodic follow-up, which remains under the responsibility of the utility.

In all cases, specific syntheses have been prepared to summarise the knowledge gathered on thevarious topics, and the situation of the various units and components in front of the ageingphenomena:

• In Belgium, equipment ageing summaries are established with priorities based on importancefor safety and availability, following a defined format in order to provide a status on activitiesundertaken.

• In France, due to the standardisation of plants, generic studies have been done on Plant Life,with detailed reports on main systems and components, and on various ageing mechanisms.

Where needed, specific projects have been established for a given period on major ageingphenomena, the results obtained being then included in the plant management practices.

• In Spain, component degradation data sheets and maintenance practice data sheets are preparedto facilitate the systematic evaluation of existing practices and the eventual improvementproposals.

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There is a common approach to prioritise components, according to:

109

1. The evaluation of the safety risks, which depend on the potential degradation phenomena andtheir associated risk of component failure, and on the potential consequences of such a failure,

2. The difficulty or cost for component repair or replacement. Some components may determinethe life of the plant itself, such as the reactor pressure vessel or the containment.

The procedure is more particularly developed in this report for the Spanish practice, butcorresponds to the general approach of the two other countries as far as the technical parameters areconcerned.

Consequently, there is a general agreement on the main classification of important componentsobtained according to these methodologies, with non-replaceable components at the top anddifferences on classification details which are essentially the result of differences on utilitiesstrategies, and the corresponding weight given to the different aspects.

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The identification and selection of ageing mechanisms is based in every case on systematicprocedures which serves as the basis for the orientations given to the in-service surveillance andmaintenance. This evaluation is further enriched by the consideration of the return of experience,including the evaluation of international information.

The first in-service inspection programs were based on design assumptions and engineeringjudgement, quickly completed by ageing experience, which generally appears where notanticipated. There are good reasons for that: where ageing phenomena are anticipated, designchoices, including system design, component drawing, material choice, tend to provide adequateresistance for the expected component life.

The practical organisations for ageing mechanisms identification in the various countries fordifferent components are consistent, with the following specificities:

• Belgium retains a systematic structure (given in Appendix 3) for components important from asafety point of view. Where several ageing mechanisms are identified for a given component,there are several EAS established accordingly.

• France establishes synthesis documents on component life – component by component – on onehand, and on the most important degradation mechanisms covering their effect on variouscomponents on the other hand.

• Spain follows a systematic process for the evaluation of ageing problems and correspondingcorrective measures, establishing the adequate pairs of SSC and ageing mechanisms /degradations / mitigations.

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The presentation of practices on typical ageing mechanisms is given in Appendix 3. Illustratedexamples common to all three utilities are discussed below. They cover:

• neutron irradiation embrittlement (of reactor pressure vessel),• thermal ageing (of austenitic-ferritic castings essentially),• stress-corrosion cracking (particularly of 600 alloys),• fatigue,• reactor building ageing,• electrical components and cables ageing.

Differences between practices are explicitly identified. Where no reference to a particular country ismade, the corresponding practices are considered equivalent.

�� 1HXWURQ�IOXHQFH�HPEULWWOHPHQW

110

The potential effect of neutron fluence on fast fracture risk of the reactor pressure vessel isevaluated in all three countries. The various aspects of this ageing mechanisms may be compared asfollows:

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In all cases, the so-called "Belt Line" is considered for fast fracture evaluation. This zone may bedefined with respect to the neutron fluence likely to be calculated or observed, or on the expectedeffect (for example, a minimum shift in transition temperature).

The second definition was first used in France, which did lead to necessary evolutions and in somecase difficulties when the knowledge of ageing behaviour evolved. This did lead to an evolutiontowards an irradiated zone defined as the zone subjected to a fluence greater than a given threshold.

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The effects of neutron fluence include a shift of the brittle-ductile transition temperature, as well asa decrease of the upper shelf toughness. All practices consider both effects, the second one beingmore explicitly considered where fast fracture prevention is not limited to non-ductile (brittle)failure risk, but also includes ductile tearing evaluation, as it is the case in France.

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Prediction of neutron fluence effects is obtained using US-NRC Reg. Guide 1.99, rev.1 [103](included in RCC-M Appendix ZG) in France and rev.2 [104] in Belgium (for old plants) andSpain. French practice do not use rev.2 due to the fact that this formula is based on in-servicesurveillance of US materials not representative of materials with low residuals. This is the reasonwhy more precise in-service evaluations are done using FIM and FIS formulas, based on dataobtained on representative materials and codified in the RSE-M code. Belgium EAS recognises thisfact by using FIS and FIM prediction formulas for more recent units.

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Surveillance programs are defined according to rules generally based on US practices. Specimensrepresentative of base materials, welds and heat-affected zones taken from the belt line region areincluded in capsules subjected to a neutron flux higher than the one applied to the RPV wall itself,leading to a given "anticipation factor".

Two considerations may be applied here:

• requirements on the maximum anticipation factor in order to guarantee the representativity ofageing measured on test samples, like in France where the fluence on test specimens shall belower than three time the one received by the RPV wall,

• or requirements maximising the anticipation factor, leading to a minimum value for this factor(> 3 in Germany, for example).

These capsules are withdrawn according to a pre-determined schedule in each country. There is aglobal consistency of practices on this topic.

The embrittlement is generally monitored by measuring the transition temperature shift at aconventional Charpy V toughness level of 41 J. The material toughness curve is then shifted by thesame value.

There is nevertheless a general tendency towards including fast fracture specimens in thesurveillance program, allowing a direct determination of the applicable toughness curve of thematerial, and a reduction of uncertainties, or instrumenting Charpy tests. Evaluation of Chooz Aissue in France and of Doel 1&2 in Belgium particularly demonstrates the pessimism of ageingpredictions based on the above conventional approach.

In addition, reserve capsules are provided with the objective to cover life extension needs, inparticular in France and in Belgium.

111

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For normal operation, two sets of criteria are used:

• The US approach used in Belgium and Spain, where a reference conventional defect isconsidered and a safety margin is applied on the part of the stress intensity factor which is dueto primary loads (mainly due to pressure),

• and the French approach, which considers more realistic defects and margins applied to not onlythe primary part, but also to the secondary part. France also consider criteria for brittle failureprevention and ductile tearing. Partial safety coefficients are also codified in the RSE-M, basedon semi-probabilistic considerations.

Also screening criteria to limit the risk of vessel failure due to pressurised thermal shocks areapplied in countries whose practices are based on USA approach. These criteria, based on values ofRTNDT at end-of-life, were prepared, but not accepted in France. Demonstrations shall cover thenext ten-years interval, and demonstrations covering the whole life have been prepared, but have noregulatory value.

There are no real limitations of the reactor vessel life, except if renewal of pressure tests has to beconducted at a pressure higher to the design temperature (1.20 Pc in France), with a minimummargin between test temperature and transition temperature, the test temperature being lower than100°C for safety reasons. This may lead to vessels which cannot be used for the reason they cannotbe tested safely according to regulatory requirements.

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In all cases, mitigation of neutron fluence effects may be obtained through a reduction of fluence byincreasing at design stage the vessel diameter, and by using at design and/or in-service stages a coreloading pattern with a lower leakage.

Another possibility is a vessel annealing, which is identified, but not used by the utilities of thethree countries.

The reduction of fast fracture risk may also be obtained through a reduction of thermal shocks andan increase of material temperature during an accident, which can be obtained by an increase of thetemperature of the safety injection water (used in France and Belgium). In the case of the Chooz Aissue, the criterion retained was a maximum difference between the transition temperature of thevessel material and the temperature of the safety injection system.

Fast fracture risk is also reduced through qualified non-destructive examinations demonstrating theabsence of defects or providing a good detection of potential defects in the most severe zone, i.e. the"first 30 mm" of the vessel shell, as retained in France.

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Thermal ageing may affect essentially austenitic-ferritic castings used for pump and valve casingsand cast elbows. Thermal ageing may also affect pressuriser low alloy steel through the form of asmall shift in brittle-ductile transition temperature. Thermal ageing of austenitic-ferritic castings areconsidered by all utilities.

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The surveillance concerns mostly the reactor coolant pump casing and the cast elbows. Specificfiles have been established in France for "hot" and "cold" elbows, ageing effects being stronglydependant on temperature.

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Thermal ageing leads to a reduction in material toughness (Charpy values and J tearing resistanceproperties). Practices in the various countries have a similar coverage.

112

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Prediction of ageing effects have been determined, based on temperature, chemical composition andequivalent ferrite content and on models validated on experiments. French and Belgium provisionsare very similar.

French studies are based in particular on a very important program including tests and calculations,justified due to the standardisation of French plants, which has no equivalent in the other countries.


Belgium and Spanish practices follow US practices. Belgium uses Code Case N-481 allowing thereplacement of volumetric examinations, subject to material and fast fracture evaluation conditions,for the primary pump casing. French practice is described in the RSE-M, with additional specificprovisions coming from the conclusions of the generic studies for the most affected components.

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There are no specific criteria in Belgium and Spain on thermal ageing of cast stainless steelproducts. A leak before break evaluation may be acceptable.

In France, margins against fast fracture have to be demonstrated, taking into account the defectslikely to affect the components. Cast elbows replacement need shall more closely be evaluatedwhen the next steam generator is replaced, when the equivalent ferrite content is larger than 25.5%.A LBB evaluation was only used for a complementary technical evaluation according to a "defence-in-depth" approach, but not used as a safety criterion.

113

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There are no practical mitigation possibilities, which would reduce material ageing. The accidenthypotheses remaining conventional, the only possibility to reduce the fast fracture risk is byreducing the defect probability through non destructive examinations, and by replacing the mostaged components taking the occasion of the SG replacement.

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Stress corrosion cracking on 600 alloys is an important issue, considered in all countries.

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Zones covered include all zones important for safety where return of experience has shown SCCrisk, including in particular SG tubes and RPV head penetrations.

Depending on the strategy used for prediction of ageing effects and surveillance programs, otherzones which are less important for safety but which are considered "precursors" are subjected to in-service surveillance in France. This surveillance is part of a global surveillance of all Inconel zones.

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Stress corrosion cracking leads to a risk of crack initiation, which after propagation may lead to aleak, and in some cases to a failure risk.

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Different predictive models are established by the various utilities, based on statistical evaluation ofhistorical results.


US-based (ASME XI) inspections are applied in Belgium and Spain. In France, the minimumrequirements are included in the RSE-M. Complementary programs are applied in each countrywith a percentage of examination and a periodicity which depend on the prediction made for theunit/component.

Eddy Current is generally used for SG tubes inspection. Leaktightness tests are also applied, andon-line monitoring of primary/secondary leaks through activity measurements allows following theevolution of an eventual leak.

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Criteria on SG tube cracks include a percentage of the thickness (40% in general) and a length, bothlinked to rupture risks evaluation.

Criteria on vessel penetrations are linked to the prevention of leak risks, taking into account thecrack propagation risks during the next inspection interval. Tube plugging criteria are linked to SGrequired performance. The general acceptable percentage is 10%, but may be increased subject tojustification.

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Methods for mitigation of stress corrosion cracking phenomena include an increase of initiationtime by a reduction of temperature (vessel heads), a material heat treatment reducing SCCsensitivity, an improvement of water chemistry, and surface treatments creating compressivesurface stresses, and last but not least component change. This last solutions was considered moreeconomic for vessels heads in France. Steam Generators are also changed when performances dueto tube plugging are no more compatible with plant requests.

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114

Fatigue evaluations are conducted at design stage, leading to a usage factor which shall remainlower than 1.0 at end of life. Transient follow is done during the life of the plant to demonstrate thatservice conditions are consistent with design assumptions. This practice is common to all countries.

Consequently, return of experience is essentially limited to components subjected to complexthermal loading, which were underestimated at design stage, and to components for which adetailed fatigue evaluation was not required due to their safety classification.

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Fatigue initiation under thermal loading is considered in zones where return of experience did showpotential initiation risks, for example thermal barrier of reactor cooling pump, or zones with mixingof fluids at different temperatures.

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The final objective of fatigue prevention is the failure or leak risk. Every country agrees on thisobjective. Nevertheless, the best way to reduce fatigue risks is to reduce fatigue crack initiationrisks. There is a difference on the importance given to this topic. Prevention of initiation is morestrongly required by the safety authority in France, leading to practical criteria more stringent thanUS-based criteria

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Prediction of fatigue risks is conducted through fatigue analyses using standard procedures. Specificevaluations are applied where needed in zones with stratification or complex thermal loading.

Evaluations of crack propagation risks in thermal barriers show that the number of cycles necessaryfor having a failure or even a leak risk is high and is compatible with optimised replacementprogram.


Transient monitoring is applied in all countries, though through slightly different practicalimplementations in order to evaluate applied transients and their numbers of occurrences.

Periodic inspections are conducted according to ASME XI in Belgium and Spain, and according toRSE-M in France.

No systematic surveillance programs are required for thermal barriers of primary pumps.

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Prevention of fatigue is obtained through ASME III criteria in Belgium and Spain, and throughsimilar RCC-M criteria in France. Initiation prevention is required in France. Prevention of leak orfailure risk is required in all countries.

Evolutions are under evaluation in the various countries in order to extend fatigue verificationrequirements to old class 1 piping not initially subjected to fatigue evaluation requirements, and tonon-class 1 components subjected to severe loading, and to reevaluate the applicable fatigue designcurves.

115

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Mitigation of thermal fatigue risks include recommendations on system use (residual heat removalsystem case), the suppression of leak sources, improvement of design details and surface conditionin case of component replacement. A particular attention is paid to this last consideration in France.

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Containment functions covered in all countries include resistance and leaktightness. Resistance isaffected by decrease of tension in prestressing cables. Ageing of concrete may lead to potential lossof leaktightness. Relative movements between structures are also addressed.

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Requirements cover essentially concrete buildings and particularly containment buildings with nolining for leaktightness aspects.

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Loss of tension of prestressing cables, cracks leading to leaks, relative movements.

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Containment evaluation is mainly based on kinetics derived from actual tests.


Surveillance programs include equivalent provisions in the various countries, covering:

• strain and deformation measurements of concrete walls,• tension tests of cables, and inspection of anchors,• absolute and differential settlement measurements,• visual examinations,• containment pressurisation tests and leaktightness tests• mechanical and chemical tests are also conducted on test samples (Belgium).

Frequency of inspection depend on the country and on the test under consideration, but remainrelatively similar.

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Criteria include:

• minimal mechanical properties of the structure,• design settlement values,• leaktightness requirements, expressed as an allowable percentage of internal volume per day.

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Mitigation methods include repair of concrete and injections.

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Electrical equipment classified 1E is qualified according to provisions equivalent in variouscountries, which shall include ageing risks (VISA procedure according to French RCC-E, US basedpractice in Belgium and Spain). IEEE standards are referred to in all rules.

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The objective is in all cases to be able to demonstrate the validity of the qualification tests.

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116

This is obtained in all cases by the simulation of the significant parameters (ageing, irradiation,earthquake and simulated accident conditions) for the intended qualified life, which are appliedduring the qualification tests.

In addition, a large generic study was conducted in France in the context of the plant life evaluationstudies.


Qualification of components is covered by qualification programs.

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Components have to be replaced when qualified life expires. Extension of qualified life may bepossible subject to the verification that components have been subjected to less severe conditionsthan expected.

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By preventive maintenance / replacement programs. Obsolescence have to be particularlyanticipated for these components. This is in particular the task of a large French study involvingvarious potential suppliers.

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Ageing of insulation is anticipated in qualification tests, except for older cables, where such testswere not done. All countries appear to conduct a specific evaluation of cable ageing and associatedrisk of loss of electric properties likely to affect their intended functions.

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Loss of electrical isolation due to thermal ageing and irradiation and risk of loss of function of K1classified cables during accident. This concerns mainly PVC insulation. Reduction of strain atrupture for all materials.

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Anticipated ageing conditions are covered in the qualification tests conducted on representativecomponents.

Models have been established, based on collected experimental data, including internationalreferences and laboratory tests.


Monitoring of mechanical and electrical characteristics for original non-qualified cables + routinevisual examinations or monitoring of representative samples, depending on condition severity.

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Cables are replaced before the end of their qualified life.

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Extension of life is possible through a better evaluation of the environment severity.

117

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The above comparison conducted on some typical examples show the global consistency of theapproaches chosen in the three countries, though the applicable criteria may differ on particularpoints, such as criteria or prescribed safety margins.

The general tendency is for Belgium and Spain to refer as far as possible to the US methodologyand criteria, where France tends to develop its own methodology due to the standardisation of itsNPPs and specific regulatory requirements.

118

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�� )25(:25'The purpose of this analysis is to compare the approaches adopted by 3 countries, Spain, Belgiumand France for an Ageing Management Process, with the international recommendations.Additional comparisons with US methodologies are given in Appendix 5.

The methodology followed for this comparison was as follows:

1) a description by each country representative of its own process, covered in Appendix 4,2) a general comparison that constitutes the purpose of this Section.

�� '(6&5,37,21�2)�$3352$&+(6�%<�&28175<In the interests of obtaining comparable descriptions, a standardised form of questioning wassought: the reference document for the IAEA Ageing Management Assessment Teams (AMATsdated March 1999 [44]) includes in part II (Supplementary Guidance for AMP programmaticreview) a list of areas to be reviewed: as a reminder these are given in Table 6.1 below.

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- Regulatory policyand requirements

- AMP policy

- Internationalguidance

- Scope of AMP

- AMP organisationand programmedescription

- Resources:

(a) human

(b) financial

(c) tools andequipment

(d) external

- Provisions forunderstanding SSCageing

- SSC screeningmethod

- List of SSCs

- Operationalprocedures

- Surveillance

- Assessment

- Maintenance

- Data collectionand record keeping

NB: assessmentincludes dataanalysis

- Physical conditionof SSC

- EQ established andmaintained

- Performanceindicators

- Self-assessmentprogramme

- Peer reviews

- Comprehensivereviews

- Continuousimprovementprocess

The 3 countries filled in a form based on these areas, except for II.4 "results" which was notrelevant. The completed forms are given in Appendix 4:

− Spain: completed by UNESA,− Belgium: completed by Tractebel,− France: completed by EDF

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119

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Each country have defined or refer to existing documents on the following topics:

− Regulatory policy and requirements, covered in Section 4,

− Ageing Management Processes covered in Section 5, described in UNESA documents in Spain,in the context of the "Continuous Operation of Belgian NPPs" project in Belgium, and by the"Service Life and Ageing of Pressurised Water Power Plants" project in France,

− International guidance documents are considered in each case, these documents includingcontributions of contributors to the present study.

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The Safety and Regulatory aspects are covered in Section 4. The life time of the units in the threecountries is not limited by regulations:

- Spanish and Belgian units are subject to a licensing process,- French units obtain their authorisation to operate through a «décret de création».

In no case is a limit pre-determined and, provided that the safety level can be justified, the decisionto continue to operate is taken by the Owner on economical grounds.

The safety status of the units is subject to a continuous process of assessment in order to validate theconformity to the Safety Analysis Report. In addition Periodic Safety Reviews (P.S.R.) areperformed and may include a Safety Reassessment.

- a P.S.R. is required under the terms of the licence for Spanish and Belgian units,

- in France, EDF has proposed and the Safety Authority has accepted to perform the review ofconformity and the reassessment on a 10 year basis.

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With different levels of formalisation of criteria (safety, availability, economical…) the principle isin every case to define a list of sensitive components that have a risk of failure due to ageing beforethe forecast life duration.

The design life duration is given by figures such as 30/40 years: it can be mentioned in the SafetyAnalysis report; the value may be different for certain equipment.

Sensitive equipment is ranked according to the degree of importance for lifetime management,taking into account the importance for safety and availability.

It appears that the Ageing Management Programs of all countries go beyond the IAEA objectives,which are limited to safety aspects, to cover all components which have a significant impact onreliability, replacement and cost.

For these components the scope of the programme includes:

- analysis of ageing phenomena,

- justification of behaviour, including rate of deterioration and determination of critical defects,

- assessment of maintenance practices.

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120

- UNESA and Tractebel do not have direct responsibility for the units: they are in charge ofmethodologies, advice, studies which should be taken up and put into force by the managementof the units.

- EDF has to manage series of units, so it imposes practices on all units of the same type from acorporate level.

The consequences:

• For Spain: each operator has its own organisation, which has to adapt the programme proposalsand the results of the studies carried out by UNESA to its specific circumstances. Responsibilityfor discussion with the Safety Authorities is in the hands of each unit.

• For Belgium: the basic rules are given by Electrabel at a corporate level and then, on each sitethey are put into action by a local organisation. The engineering support is given by Tractebel.The responsibility for discussion with the Safety Authorities is in the hands of each site (i.e.Doel or Tihange).

• For France: EDF manages the programme at a corporate level of the Operating Division (DPN).On request, the EDF Engineering Team and Research Department provide the supportingactivity. The responsibility for discussion with the Safety Authorities is in the hands of thecorporate level of DPN.

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In the three companies the activity is managed by specific teams:

- in Tractebel and UNESA they are project type organisations. They submit their product to theirrespective clients, Electrabel or the Spanish Units, for decision and implementation.

- in EDF: it is a co-ordinating group, which puts recommendations to the Steering Committee incharge of making all decisions for modifications or maintenance (either basic design or detaileddesign and implementation of modifications).

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The personnel in charge of the activities linked to lifespan, either directly involved in the specificteams or maintained in their original entities, have not been chosen according to specific criteria: asis the case for any other activity, they are chosen for their skill and experience in relation to the taskassigned to the team they belong to.

Life management in fact demands a wide variety of skills and it appears clearly that no dedicatedorganisation has been set up in the engineering teams. The specific team in charge is thus requiredto co-ordinate a large number of actions assigned to specialised teams.

As a result no personnel is allocated; execution depends on the capacity of the teams, laboratoriesand subcontractors to handle the workload of the job assigned to them. As is the case in anyengineering job, they have to anticipate and forecast their activity. Accordingly, training forpersonnel is decided in observance of the internal procedures of the entity.

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In addition to internal resources, each entity can decide to complement the skills needed by externalsupport from specialised companies. Exchanges between utilities and engineering companies arenumerous and organised. Periodic meetings or congresses are held within the scope of the EC,IAEA and WANO.

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121

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Belgium, French and Spanish methodologies may be summarised as follows. These descriptionsshow that IAEA recommendations are fulfilled in every case, although the practices described gobeyond IAEA methodology, which is limited to safety aspects.

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The methodology defined in IAEA Technical Report Series N° 338 [37] is oriented to ensuringsafety during the service life of the plant by knowing, controlling and monitoring ageingmechanisms that could put to risk plant safety.

The methodology is structured as follows:

a. Selection of components important for safety and whose ageing must be analysed. Theselection process includes the following steps:

1. Selection of the systems and structures that contribute to plant safety2. Impact of component failure on system functions3. Probability that ageing can cause component failure4. Suitability of maintenance

b. Methods to analyse said ageing effects, as well as the options for monitoring and mitigation.

Initial selection includes all SSC that are safety-related and those that are not safety-related andwhose failure can prevent performance of safety functions. However, strict and systematic ageingassessment is only applied to the population which, exclusively in accordance to the experiencegained by the industry and to the plant specific experience, are subject to failure by ageing and arenot covered by appropriate maintenance for ageing mitigation and/or monitoring

The purpose of ageing assessments is to detect ageing mechanisms that could prevent safetyfunction performance in the course of the plant service life period

The assessments are aimed at determining the severe effects of ageing on components andstructures, and at confirming maintenance efficiency to mitigate andlor monitor said effects,keeping components within safety margins for them to fulfil their functions during service life.

Lack of sufficient information or of precise knowledge of degradation mechanisms that generateuncertainties in the initial "provisional" assessment (phase I) of ageing effects require detailedanalyses (phase II) to clear said uncertainties.

Maintenance evaluation is performed with criteria similar to the ones in the other methodologies,but for the service life scenario.

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�� 6SDQLVK�0HWKRGRORJ\

The UNESA methodology comprises the following steps:

a. Selection of components, using life management criteria,

b. Examination of ageing mechanisms and selection of components subject to severe degradation,

c. Evaluation of maintenance for life management and proposal of improvements.

Initial selection includes safety-related SSC as well as those that have significant impact onavailability, replacement and cost.

The entire population selected is submitted to systematic ageing assessment; when ageing effectsare severe, maintenance efficiency is assessed.

The purpose of ageing assessments is to detect ageing mechanisms that could prevent performanceof the functions necessary for safe and economic operation during plant service life.

The assessments are aimed at determining ageing effects on components and structures, and atidentifying the ones that are severely affected and therefore require a detailed analysis ofmaintenance practices to confirm ageing control during service life.

The entire process is summarised in the component degradation data sheets and in the maintenancepractice data sheets that are prepared to facilitate the systematic evaluation of existing practiceswith the subsequent improvement proposal.

Maintenance evaluation is performed with criteria similar to the ones in the other methodologies,but for the service life scenario.

�� %HOJLXP�0HWKRGRORJ\

In Belgium, the safety of nuclear power plants is based on compliance with fundamental safetyprinciples which must be applied in everyday operation, following lessons learned from nuclearincidents, and on a regular and comprehensive review of safety in order to ensure that any necessaryimprovements are made. During safety reviews, ageing is systematically investigated in order todemonstrate that the safety of the installations is guaranteed during the next decade.

For safety-related components, specific programs exist in order to guarantee the integrity and theavailability of these equipments during the exploitation of the NPPs.

For the passive safety-related components and the non safety-related components but important forthe plant availability, a specific ageing management project was created. The main objective of thisproject is to centralise all safety and economic aspects of the plant life management in order todetermine, for each unit, the most probable cost required to maintain, safely and economically, theunit in operation.

The selection and prioritisation of SSC is performed with criteria similar to the ones in othermethodologies (e.g. IAEA), although the IAEA methodology concentrate only on safety aspects.

In order to facilitate ageing assessment, Equipment Ageing Summaries were created. These reportssummarise the knowledge gathered on the various topics related to SSC ageing and the situation ofeach NPP with respect to these ageing phenomena.

�� )UHQFK�0HWKRGRORJ\

EDF Methodology is the following:

1. The purpose of ageing assessments covers three aspects closely interconnected: technical,economic and safety

123

7HFKQLFDO: equipment life may be jeopardized because of various deterioration modes inhibitedrequired functions. Component ageing depends on operating and maintenance conditions.Therefore solving technical issues requires taking into account the following aspects : designand manufacturing file, operating conditions, maintenance strategy, experience feedback andsupport from R&D programmes.

(FRQRPLF: the cost of nuclear power generation needs to remain competitive while integratingmaintenance costs involved in dealing with ageing effects and those induced by compliancewith increasing safety requirements. For example this means that it is important to correctlyassess probable trends in "standard" and "exceptional" maintenance costs.

6DIHW\: the operation of installation needs to always comply with regulations. Ageing effectsmay reduce design-basis safety margins wich may in turn generate changes with operatingconditions. These margins may decrease due to increased safety requirements.

2. The assessments are aimed at determining the potential life of components and structures bytaking into account all available aspects (see "technical" in the above paragraph ). EDF givespriority to routine and exceptional preventive maintenance activities because both imply a so-called "anticipation approach". The objective is to identify and assess generic-type defects ofcomponents and systems. This relates to functional faults or degradation, component ageing,and manufacturer diminished skill.

Exceptional maintenance involves sporadic though possibly generic activities, usuallyperformed once in a plant life and involves significant resources and cost.

Beyond the analysis of equipment ageing, overall generic approaches are being developed,focusing on specific generic-type damage : irradiation/embrittlement, vibration and wear,corrosion/erosion, non ductile failure, stress corrosion, and non destructive measurement ofmaterial degradation (through ageing).

�� 23(5$7,1*�352&('85(6

Operating procedures are written in compliance with design data, such as the number of transientsthat items of equipment subject to fatigue analysis are to undergo.

Further provisions have been added so as to limit ageing effects:

- limitation in terms of the profile of transients and their occurrence,- improvement of secondary side chemistry,- arrangement of core for low neutron losses...

A detailed list of these provisions has not been made (an analysis unit by unit would be necessary)but it is understood that the subject is handled by the three engineering organisations and thecorresponding utilities.

Of the topics checked through the list in the AMAT guide, this one is particular in that it aims toavoid ageing. All the other topics, without exception, try to detect and correct the effects of ageing.

�� 6859(,//$1&(�352*5$00(

They exist in the three countries: include periodic tests and surveillance programme.

The feeling is that they have been issued independently of the ageing programme. However, afterspecific reviews as explained for example by UNESA, it may be decided to complement them. InFrance the «preventive maintenance basic programmes» were decided subsequent to analysis of thefailure rates. They are complemented by the feedback from experience on incidents (specific checkpoint added) and progressively optimised to reduce cost – a major factor with regard to life time –.

�� 0$,17$,1,1*�(48,30(17�48$/,),&$7,21

124

The life time of qualified components is determined by the hypotheses adopted in their qualificationprogramme:

- for components subjected to fatigue, reassessment is under way (real number of transientoccurrences, less conservative methods of fatigue analysis),

- real conditions are also checked against qualification programme analysis,

UNESA / NPPs focus on the problem of obtaining spare parts, as many suppliers areabandoning the production of nuclear-grade components.

EDF is adding a process to inform the maintenance team of particular points needing careduring maintenance operations, so as to avoid impairing qualification.

�� &203$5,621�,1�7(506�2)�021,725,1*The programmes have been presented to the Safety Authorities, which have given a favourableopinion.

The results are included in the yearly review and, as such, are subject to review by the owner andsubmitted to the Safety Authorities for approval.

In France specific "Groupe Permanent" meetings are held for this purpose.

�� &21&/86,21This analysis shows that:

• ageing management process is a huge activity in the three countries.

• No legal life time is imposed. Only design life times are mentioned for some components andare the starting point for analysis.

• specific teams have been created to manage the process, with either a project type or missionco-ordination type organisation.

• the engineering job is assigned to the teams according to their skills: no devoted entity has beenestablished. The teams called upon may be from in-house engineering staff or subcontractors.For the moment, there is no apparent difficulty in modifying the resources.

Only in some cases do difficulties arise in terms of spare part availability or engineering, as aresult of companies abandoning a certain field.

• Responsibility is always borne by the operator: by the units in Spain, at a corporate level byElectrabel and EDF in Belgium and France respectively.

• The involvement of Safety Authorities is always strong:

− presentations of the programme are made,

− through PSR, ten-yearly assessments are performed and formalised.

• The working method is more or less formalised, but always includes:

− selection of sensitive equipment,

− identification of ageing phenomena,

− cross analysis with the maintenance and monitoring processes applied to the selectedequipment in the units.

• Two levels of activity appear:

− major components (reactor vessel, primary loops, containment) for which in-depth analysesare necessary to justify life duration,

125

− smaller components for which analyses are required to detect and anticipate failure, in whichcase the links with planned surveillance programmes and preventive maintenanceprogrammes are established.

• The aim of the utilities of the three countries is to operate their units as long as they can operatesafely and economically, with the support of their own engineering forces and subcontractors. Inthis analysis, the question as to how to maintain the necessary skills over such a long period, ascompared to other industrial processes, is not included. The communication of the process tomaintain the confidence of the population should also be part of ageing management.

126

�� &21&/86,216�$1'�5(&200(1'$7,216

�� &21&/86,216Safe control of ageing of nuclear power plants constitute an important concern for every plantOwner and Safety Authority. Since 20 years, this topic constitutes the subject of numerous studieswhich have led in particular the utilities to establish programs or projects specifically dedicated tothe management of ageing of systems, structures and components (SSC).

The study shows that Belgian, French and Spanish nuclear power plants are subjected to appropriateageing monitoring programs or projects, which fulfil the IAEA international recommendations onthis topic. The contributors of the present study agree on the conclusions of the international groupsand consider that these recommendations are sufficient to help operators managing ageing aspectsfrom a safety point of view.

The study also shows the general consistency of the objectives in the different countries, which gobeyond safety aspects, differences in organisational approaches resulting from differences inindustrial and regulatory contexts. There is in particular a general agreement on the service life ofthe plant which is the life during which the plant may be operated safely and economically. Safetyis then a necessary, but not a sufficient condition.

Current ageing management programs aim essentially at managing the gradual evolution ofsystems, structures and components as a result of their physical ageing in order to ensurepermanently satisfying the safety criteria. Apart from this physical ageing, other factors will sooneror later affect plant service, such as climate changes, evolution of electricity market competition andindustrial context, organisation changes, conservation of human knowledge, evolution of safetycriteria, management of human factor on a long period... Such parameters only start beingconsidered in ageing studies.

The various aspects leading to slow evolutions (management of physical ageing and generalchanges) are in particular evaluated during periodic safety reassessment (generally each ten years).Aspects related to more quick changes (in particular those affecting active components) aremanaged on a continuous basis, through an appropriate maintenance, and component qualification.

It is possible to conclude that the "ageing" of LWR nuclear power plants can be put under control.Through monitoring and clear understanding of the different degradation mechanisms, which arecommon phenomena in an industrial facility, the operator is able to anticipate, via an adaptedprogramme, the necessary measures (for instance, monitoring, servicing or replacement of materialinvolved) in order to operate its plants safely and economically.

�� 5(&200(1'$7,216Considering the existing practices, the results obtained and the research programmes currentlyrunning, the ageing management field may be considered as being adequately covered.Nevertheless, one can suggest to particularly emphasise the efforts on the following topics:

127

Regulatory aspects

Electricity operators have more and more to compete in a de-regulated market with other countriesand other energy sources and they are subjected in the nuclear field to national requirements whichmay be significantly different on some aspects.

It is therefore recommended that exchanges between regulators be maintained at the European andInternational levels, and that harmonisation of general provisions be obtained in safety standards,particularly on the consideration of risk-informed aspects, which constitute one of the essentialbases of any plant ageing management approach, and qualification methods.

In the context of operator competition, it is also recommended to agree as far as possible, throughEuropean or international safety groups, on the expression of safety acceptance criteria (includingprobability figures, and/or safety margins objectives, where appropriate).

Although a permanent attention to safety aspects is necessary, a certain stability of the safetyrequirements is recommended to permit full benefits from the analysis of return of experience andplant improvements.

Management aspects of physical ageing

Following the conclusions of the present study, it does not appear necessary to formulate particularrecommendations related to ageing management of plants in general. It appears nevertheless usefulfor operators to issue periodic syntheses following IAEA format. The existence of such syntheticperiodically updated evaluation may also play a role in the general context of public acceptance.

In particular, it is recommended to consider the logical Deming’s Plan-Do-Check-Act approachdescribed in [38] (See Appendix 1, B4 sheet).

A particular emphasis shall be placed on a close follow up of what happens in the plant. In additionto IAEA reports recommendations concerning design information, the weight shall be put ondetailed operating conditions book-keeping covering the entire life of SSCs.

Ageing management shall not be restricted to the oldest plants, where improvements or mitigationmeasures may have a limited impact. It shall consequently be considered as early as possible in thedaily operation of the plant.

Prediction of ageing

Damage can never be 100% anticipated. Design studies can only restrict (though significantly) theuncertainties. From this point of view, the developments concerning identified ageing mechanismsare considered adequate. The essential points where progresses are recommended concern:

− the evaluation of degradation kinetics, in order to avoid failures occurring earlier than expected,

− the description of uncertainties, problems of ageing evaluations being related to the accuracy ofthe description of the imposed loading or environment conditions,

− the improvement of the understanding of local loads and their variations as a function ofoperating modes, including the starting tests and eventual modifications, which shall beconsidered as part of the loading history,

− a better evaluation of the effect of fabrication process on ageing (for example surfaceoptimisation intended to modify residual stress field), and the improvement of the knowledge onsurface behaviour. As far as the majority of in-service damages appear at the componentsurfaces, it is useful to give a great attention to surface engineering, by exchanging knowledgebetween experts of various disciplines, such as chemists for the behaviour of oxide layers,manufacturers for the control of surface state and residual stresses, thermal-hydraulic experts for

128

the determination of thermal loading, mechanical analysis experts for the determination ofsystem and component behaviour...,

− the use of expertise conducted on decommissioned plants or replaced components in order toextend the available data and provide a better evaluation of existing safety margins,

− the integration of additional material where material ageing programs are conducted, to coverlife extension or evolutions in ageing evaluation methods,

Detection / follow of physical ageing

Detection and knowledge of physical ageing may be improved through:

− a reinforcement of cooperation between plant operators in order to identify as soon as possibleany signs of ageing precursors and their possible treatment. This does not necessarily signify theintegration of every data in large databases, the practical applications to specific series beingbetter based on specific narrower distributions,

− extending relations with other industrial sectors in order to exchange experience on ageingphenomena not specifically linked to irradiation, in particular concerning civil works,

− advance tools for monitoring and surveillance in order to permit a better evaluation of existingmargins and a better allocation of available resources,

− continuing improvements of traditional non-destructive examination methods, in particular inconcrete structures, and industrialisation of methods likely to detect ageing risks beforeoccurrence of irreversible damage (during damage initiation),

Ageing mitigation and repair / replacement

Recommendations on mitigation of physical ageing concern the following aspects:

− to optimise operation and component qualification conditions following return of experience,

− to investigate the repair methods used in other industrial sectors, in order to adapt them, afterqualification, to nuclear applications, and to share qualification costs, in order to optimise inparticular the behaviour of surfaces, the importance of which was recalled in above paragraphs,

− to anticipate replacement where appropriate, through feasibility studies including considerationof potential impacts on safety,

− more generally, to support exchanges between users and manufacturers in order to betterappreciate the relations between operation conditions and ageing.

Concerning component replacement, additional work on criteria for possible equivalence (forexample for I&C) to cover technical obsolescence is recommended.

Other aspects of ageing management

In addition to the consideration of technological aspects dealt with above, it is recommended toassess the management of the other aspects of ageing, which seem to present less acute directconsequences than physical ageing, but must nevertheless be evaluated. On these topics, theexperience of other industrial sectors may give valuable information, even if the service lifeenvisaged for nuclear power plants is longer than the design life of the major part of the otherindustrial installations.

In particular, a significant part of recorded incidents are due to human factors. It is thereforeessential to invest on human management aspects, and to the adaptation of safety culture andprocedures, taking into account the evolution of people experience, of tools and of general industrialculture, which give more and more importance to economical aspects.

129

Taking into account the objective of safe management during long periods, and the necessity todemonstrate compliance as a minimum to the initial safety requirements, it is also necessary to keepduring a long period, not only the basic requirements, but also a good understanding of what isbehind. Hence, the knowledge management is also an important point in the nuclear field.

Economical considerations

Production costs are related to plant availability factor and lifetime. Consequently, ageinganticipation and lifetime management are important from an economic point of view.

Taking into account the necessity for competitivity, which is introduced by the international andEuropean agreements on competition rules, while maintaining or increasing the safety of the plantsand the safeguard of investments already done, there is a strong need for a sharing of costs for theabove developments in the ageing management domain.

Even under strong economic pressures, safety must continue to be the first priority in themanagement of ageing NPPs.

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�� 5()(5(1&(6

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[82] S. Crutzen, B. Hemsworth, K. Kussmaul, M. Davies, P. Lemaitre, R. Hurst, U. von Estorff"The European Networks: NESC, AMES, ENIQ". Paper 95/701935.

[83] J. Guinovart, A. Placenti, L. Villanueva, "Overview on pre-harmonisation studies conductedby the working group on codes and standards", Smirt 14, paper Nr.515 Lyon, August 1997.

[84] S. Crutzen, R. Nichols "The effectiveness and reliability of in-service inspection", Int. Conf.under the auspice of the Int. Institute of welding, Glasgow, August 1993.

[85] H. Schultz "Overview of the CSNI Project FALSIRE", IAEA/CSNI Specialists meeting onFract. Mech. verification by large scale testing. ORNL, October 1992.

[86] AMES Steering Committee "Important Items of Ageing Research". Revision 1. November 20,1998.

[87] T. Planman, R. Pelli, K. Törrönen, "Irradiation embrittlement mitigation" AMES Report EUR16072 EN, 1995.

[88] R. Pelli, K. Törrönen "State of the art on thermal annealing" AMES Report EUR 16278 EN.1995.

[89] J.M. Grandemange, C. Pichon "Les modes de dégradation des composants du circuit primairedes REP" Journée SFEN du 8 décembre 1998 "Sûreté des installations anciennes", Paris.

[90] Research and training programme in the field of nuclear energy. 5-th Euratom FrameworkProgram 1998 to 2002. OJ L 26, 1.2.1999.

[91] Risk-based approach for In-Service Inspection of NPP's components. Call for tenderC2/ETU/980080.

[92] G. Van Goethem "EU Research in Reactor Safety: Achievements of the 4-th and Prospects forthe 5-th Framework Programme" Introductory lecture, Eurocourse-99 on Reactor SafetyAdvanced Nuclear Reactor Design and Safety, GRS Garching/Munich, 17-21 May 1999.

[93] "Evaluation of nuclear expertise in Europe: Situation, trends and prospects" EC Study Report,1999.

[94] RCC-M "Design and Construction Rules for Mechanical Components of PWR NuclearIslands", AFCEN, Paris, 2000 Edition..

[95] P. Lemaitre, G. Van Goethem "EU research activities in the field of plant life extension andmanagement: Achievements of FP-4 and prospects for FP-5", ASME PVP Conference,Seattle, July 23-27, 2000.

[96] Regulatory Approach to Maintaining the Safety case for Ageing NPPs. OECD/NEA, CNRA,NEA/SEN/NRA(93)10, Dec. 1993.

[97] CONTROLE - Revue de l'autorité de sûreté nucléaire. N°129, Juin 1999.

[98] INPO 97-02, "Performance, Objectives and Criteria for Operating Nuclear Electric GeneratingStations".

[99] T.D. Martin, "People Challenges at Nuclear Power Plants in the Next Century", PLIM+PLEX91 Conference Proceedings.

135

[100] CSN, Safety Guide GS-05.06, "Cualificaciones para la obtenciôn y uso de licencias depersonal de operaciôn de instalaciones radiactivas".

[101] CSN, Safety Guide GS-07.04, "Bases para la Vigilancia médica de los trabajadoresexpuestos a las radiaciones ionizantes".

[102] Y. Meyzaud, P. Soulat "In-Service aging of pressurised water reactor steam supply systemmaterials" RGN Int. Edition-Vol A- July 1996.

[103] US Regulatory Guide 1.99, Rev.1 "Effects of residual elements on predicted radiationdamage to reactor vessel materials". April 1977.

[104] US Regulatory Guide 1.99, Rev.2 "Radiation embrittlement of reactor vessel materials".May 1988.

[105] ETC-M "EPR Technical Code – Mechanical Components". Appendix ZK project. RestrictedDocument. Issue December 1998.

[106] French November 10, 1999 Order "relatif à la surveillance de l'exploitation du circuitprimaire principal et des circuits secondaires principaux des réacteurs nucléaires à eau souspression". JO. du 5 Janvier 2000.

[107] H. Churier-Bossennec "Vieillissement des produits moulés du circuit primaire principal.Coudes chauds des tranches 900 MWe. Synthèse des études". EDF restricted document E-N-M-RE-96, 1054, 1996.

[108] L. Valibus "French perspective on life management of NPP", EDF Report, March 1996.

[109] ASTM E509-86 "In-Service Annealing of Light Water Cooled Nuclear Power ReactorVessel".

[110] CSN Safety Guide GS-01.10 "Revisiones periodicas de la seguridad de las centralesnucleares, Madrid, December 1995.

[111] UNESA "Guia para desarrollar las revisiones periodicas de seguridad", Madrid, 1996.

[112] US Code of Federal Regulation, 10 CFR 50.65.

136

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137

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In the Chapter 2 "Technical issues with potential regulatory impact" the first issue presented is :ageing of plants and requests for plant life extension. Various forms of ageing are discussed, withthe corresponding regulatory challenges reproduced below :

• physical ageing of components and structures,

− to have an adequate knowledge of the current design basis of the plant,

− to have a correct picture of the actual state of the plant, through periodic tests, in serviceinspection and feedback of operating experience, in order to repair or replace aged componentsand maintain the design basis,

− to define the analyses needed to support life extensions and demonstrate that the plant will stilloperate within its design basis.

• ageing of analytical techniques and documentation,

− how to ensure that complete documentation exist to describe the current plant design,

− to make sure that safety analysis is up-to-date, reflecting the actual plant in use and allmodifications made to it,

− how to interpret results of advanced inspection techniques (are old defects being rediscovered orare they more recent ones) and what to do with defects unacceptable to modern standards,

− how to use PSA to complement the original deterministic analysis.

• ageing of rules and standards,

− applying current rules and standards to existing plants, decide which criteria should be applied,hence determining the extent of backfitting necessary. The crucial decision is defining thecriteria beyond which operation will no longer be allowed, and the difficulties involved inimplementing such criteria ; and

− checking if criteria, rules and standards developed for past technological applications remainvalid for present technology.

• ageing of technology,

− to qualify new technologies, like the use of specific software in safety critical applications or offthe shelf software for less critical ones ;

− to adapt qualification requirements without impacting on safety.

138

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The report predominantly looks at answers to a questionnaire provided by the Members countriesand the analysis performed by a task group established by the Organising Committee group.Additionally, insights from the CNRA discussions are provided in appropriate sections.

It covers following specific issues of interest for the discussion on the topic :

• the relevant safety aspects (section 4),

• the related needs for research and development (section 5),

• actions actually taken in the different countries on related issues (section 6),

• what is thought to support necessary initiatives in the safety work (section 7),

• what strategies are used (section 8),

• the matter of communicating related information to the public (section 9), and

• certain specific questions are suggested for discussion (section 10).

Among conclusions reached, the executive summary emphasises the following :

• The "broad concept" of ageing management has definite merits but remains to be generallyaccepted and understood in order to serve constructively in discussions about developments inregard of nuclear safety.

• The importance assigned to "traditional" ageing management, as seen from the national reports,and issues related to hardware degradation problems is clearly very high. The other aspects, likeengineering developments, or other types of management developments, in regard of generalprogress of the state-of-the art are considered important as well, but are less emphasised.

• Some countries see needs to develop the event reporting systems to reflect more accurately theageing problems.

• There is common interest in enhanced practices for in-service inspections and maintenance aswell as "risk informed" approaches, also as applied to selection of research projects.

• There may be reason, in regard of the need to establish safety requirements in terms of safetyupgrading in ageing management, to consider developing the use of specific criteria for thatpurpose (safety goals).

• There is some notable shift in focusing the regulatory efforts, in some countries, in the directionfrom verifying that systems and equipment meet the requirements to verifying that properorganisational arrangements are in place , that they are used in the utility processes, and thatthere is also development going on based on learning from experience. Some countries,however, continue to rely primarily on verifying the state of the plant.

• Matters concerning communicating ageing related information to the public appear to beassigned special importance in most responses. Proper arrangements seem anyhow to be inplace.

Some other interesting statements (for the purpose of the present study) are quoted below :

139

• Section 3 - The concept of ageing management

The broad concept of ageing offers advantage in bringing "management of change" into focus inaddition to the various specific issues as such which are affected by changes. No matter ifchanging conditions relate to ongoing degradation of hardware, to corresponding managementpractices, or to developing views and increased knowledge in regard of completely differentissues, the management aspects clearly constitute a common dominator.

It has been pointed out also the management of ageing, in the sense used in this report, mayrather be seen as a pure matter of quality assurance (QA). It should be observed, however, thatQA rests on quality systems, covering the relevant aspects, and that ageing management shouldessentially be seen as constituting a part of a total quality system.

• Section 4 - Safety aspects

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In nearly all responses priority is assigned at the highest level to managing the physical ageingof the plant components and structures. There are no real concerns expressed in regard of safety,however, but rather in regard of the service life of the various components and the amount ofattention and efforts required for ensuring adequate management quality.

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This concerns the gap that might be seen on viewing the current provisions for the safety andthe defence-in-depth in the light of the state-of-art. Such gap may relate to general advances invarious areas, such as the integrated treatment of man, technology, and organisation (MTO) ;safety approaches like integrated, plant specific PSA ; generally advancing safety standards andconcepts in regard of what can be considered safe enough, etc.

The importance assigned to this aspect of ageing management (management of change) isreflected in the regulatory processes, providing for comparisons to be made against modernstandards and implementation of improvements as reasonably practicable. In some countriesperiodic safety reviews (PSR) are conducted as part of the regulatory process.

• Section 5 - Research and Development needs

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Needs for further research is seen primarily with regard to ageing phenomena affecting thepressure boundary of the primary systems. Important areas include fatigue, thermal andirradiation embrittlement (and annealing), thermal shock, corrosion, erosion and cracking, andcrack initiation and propagation under the various environmental conditions prevailing and to becontrolled (e.g. chemistry) in the various primary systems, particularly in the RPV, main coolantpiping, steam generators etc ...

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A general need for research in the area of human factors and the interplay between man,technology and organisations is pointed out. Specific issues include organisational matters andinformation systems in regard of control rooms and maintenance, and conservation ofcompetence, in general and for decommissioning purposes.

140

• Section 10 suggests questions for discussions :

1. Views as to the meaning assigned to "ageing management".

2. Needs for generally enhanced assurance that the root causes of observed ageing degradationare indeed properly identified and verified and that all relevant information is secured priorto measures possibly preventing further observations.

3. Needs for enhanced feedback of information by improved reporting and recording of ageingdegradation and associated data on a broadened, possibly international basis.

4. Attention paid to inspectability and maintainability in modernisation of nuclear plants.

5. Needs and possibilities for enhanced quality assurance for improved reliability and safety ofthe nuclear plants to further ensure observation of the technical specifications of operation,e.g., in regard of ensuring operability of the safety systems after maintenance outages.

6. Need for enhanced strategies and methods in performing various types of safety review (e.g.reviews of proposed plant modifications, PSRs, re-licensing reviews etc .). - Distinctivecharacteristics of efficient PSRs.

7. Distinctive characteristics of "properly balanced" regulatory inspection and reviewing toverify, on one hand that systems and equipment meet required standards and, on the otherhand, that proper organisational arrangements are in place and used in the utility processes,including developments based on learning from experience.

8. The role of safety goals in establishing safety criteria. - is "continual improvement of safety"by itself a sensible goal in nuclear safety ?

9. Is ageing adequately addressed in PSAs regarding methods and data to support "risk-informed" approaches ?

10. Meeting the requirements of ageing management under increasing economic pressure.

141

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This workshop discussed about characterisation of mechanical properties of long time aged,irradiated materials originating from replaced, damaged, decommissioned or sampled metalliccomponents (mainly RPV, but also : core structure, cast stainless pipe, cast valve body).

It expressed the following conclusions and recommendations :

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A stage had been reached when a significant number of plants were starting to be shut down. In avery few of these cases embrittlement had been a factor in the discussion. The material from thesereactors and the associated information will be lost unless steps are taken soon.

Many material testing reactors are also being shut down. This increases the need for informationfrom irradiated material from power plant.

There were necessarily conservatisms in the arguments used in the decisions leading to plantshutdowns. It was felt that there were often significant over-conservatisms.

One of the best ways of reducing these overconservatisms in future plant shutdown decisions was totest material from shut down plant, and to compare the properties with those predicted. This was thebest way to address such issues for irradiation embrittlement as stress effects, dose rate effects, fluxattenuation effects.

Attention was often concentrated on RPV embrittlement, and the RPV was unique in the sense thatit could not be replaced, and safety cases did not include failure of the RPV. However issues such asthermal ageing, corrosion (stress corrosion cracking, environmentally assisted cracking etc) and badwelding were of as much interest to operators and regulators. In terms of components, steamgenerators, piping, RPV penetrations, valves and internals should be considered. There was also aneed for more information on defect distributions.

It was necessary to know the fabrication and service history, and the start of life properties.

The use of subsize specimens is important, and there is still a need for further development andvalidation of their use.

There were already international groupings of decommissioners (OECD-NEA Liaison Committeeand Technical Advisory Group, and CEC DG XII (Messrs Simon, Schaller). These should be usedas forum to present the case for material testing.

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Prepare a list of what was available (as in the AMES questionnaire). Such a questionnaire could bedistributed to decommissioning groups.

Prioritise this list with regard to the interests of operators and regulators and select a few items tomaximise the probability of obtaining funding.

State clearly the concerns which could only/best be effectively addressed by this means.

142

Obtain the consensus of national experts for this list through meetings such as the presentworkshop, and present the case to senior groups of scientists and regulators (CSNI, CNRA) andgroups of decommissioners (OECD-NEA Liaison Committee, CEC DG XII).

Identify appropriate sources of funding (operators/regulators for operating problems,TACIS/PHARE for VVER safety problems, BRITE/EURAM for material/scientific problems etc).

Specify clearly and succinctly to the decommissioners what was wanted, what was rejected, andhow it might be financed.

It was necessary to start with the arguments at a national level. Later the case could be made for co-operation by such means as the CEC JRC AMES network.

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This document has been established by a task group composed with representatives from Finland,France, Germany, Japan, Spain, USA.

It summarizes the long-term performance of safety related concrete structures (primarily :containments) and the national (some OECD countries not represented on the task group answered aquestionnaire : Belgium, Canada, Korea, Netherlands, Sweden, Switzerland) and internationalageing programmes in this domain (IAEA, CEC, RILEM).

A survey of structures that have exhibited degradation show that they are generally due to theeffects of environmental stressors (see table).

Primary mechanisms or factors that can produce premature deterioration of concrete structuresinclude those that impact either the concrete or reinforcing steel materials. Degradation of concretecan be caused by adverse performance of either its cement-paste matrix or aggregate materialsunder chemical or physical attack. Chemical attack may occur in several forms : efflorescence ofleaching, sulfate attack, attack by acids and bases, salt crystallization, and alkali-aggregatereactions. Physical attack mechanisms for concrete include freeze/thaw cycling, thermalexposure/thermal cycling, abrasion/erosion/cavitation, irradiation, and fatigue or vibration.Degradation of mild steel reinforcing materials can occur as a result of corrosion, irradiation,elevated temperature, or fatigue effects. Prestressing materials are susceptible to the samedegradation mechanisms as mild steel reinforcement, plus loss of prestressing force primarily due totendon relaxation and concrete creep and shrinkage.

The conclusions and recommendations are reproduced below :

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In reviewing national and international activities addressing ageing management of safety relatednuclear power plant reinforced concrete structures and liners, several conclusions can be derived :

• Performance - Performance of the structures has been good. Where problems have beenidentified, they initiated during construction (e.g., poor material quality control and prematurestripping of formwork). As these structures age, degradation due to ageing may threaten theirfitness for continued service.

• Material Data - Numerous data are available on the physical and chemical nature of concreteunder various service conditions (e.g., freeze/thaw cycling and elevated temperature). However,insufficient data exist where more than one of these ageing factors is operating at the same time.

• In-Service-Inspection Methods - When properly used and applied, in-service inspectiontechniques are effective in detecting ageing and provide vital input for assessing the structuralcondition (e.g., relevant parameters or indications of ageing processes). Techniques for in-service inspection of thick sections and inaccessible areas require development.

• In-Service Monitoring - Instrumentation systems to routinely monitor performance providevaluable data for assessing the structural condition and detecting ageing phenomena. However,these systems should be used in conjunction with periodic condition assessments.

144

• Condition Assessment - General guidance exist for conducting condition assessments. However,criteria do not exist for interpreting the data provided (i.e., a definitive answer on when ageinghas advanced to the point of degradation that requires implementation of a remedial measuresactivity).

• Repair Methods - Repair methods for general civil engineering reinforced concrete structuresare well established and effective when correctly implemented. The long-term effectiveness ordurability of remedial measures relative to their application to nuclear power plant structureshas not been established.

• Service Life Models - Service life models have been developed for estimating the remaining lifeof concrete structures. However, experience-based data are not readily available in a formsuitable for use to refine and validate the models. Applications of these models have beenprimarily to new structures with corrosion of steel reinforcement being the primary ageingfactor considered.

• Analysis methods - Several activities are ongoing that continually provide improved analysesand modelling of non-linear and time-dependent conditions. Many of these developmentsshould be applicable to analyses of nuclear power plant reinforced concrete structures.

• Structural Reliability Methods - Reliability-based methodologies provide a useful tool forquantitative assessments of current conditions and estimating future structural reliability andperformance. The methodology can be used as a basis for selecting appropriate intervals forconducting in-service inspections and determining the extent of inspection and repair activitiesto help assure continued safe operation. Quantitative data for input into the methodology arelimited and the reliability models for condition assessments have not been validated.

• Ageing Management Programmes - A number of national and international ageing managementprogrammes are presently addressing ageing. The national programmes are driven by licensingarrangements, as opposed to design life considerations, and are application specific. Someinformation exchange has initiated at the international level through efforts of organisationssuch as IAEA, but are directed mainly at structures, systems, and components other than theconcrete structures.

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Ensuring the performance and function of nuclear power plant safety-related nuclear power plantconcrete structures is important to continuing the reliable and safe production of electricity. Anumber of national and international programmes are addressing ageing issues associated with thesestructures, however, these studies are generally application specific, or in the case of internationalprogrammes, tend to be more general in that they are preparing state-of-art reports and notaddressing specific issues in depth. As a result of this, the Task Group has preparedrecommendations for a medium-to-long term CSNI programme of work.

These recommendations are :

• Holding a series of workshops that address specific issues associated with ageing. Theseworkshops have been prioritized by the Task Group as follows

½ First Priority

- Loss of prestressing force in tendons of post-tensioned concrete structures

- In-service inspection techniques for reinforced concrete structures having thick sectionsand areas not directly accessible for inspection (a round robin testing activity could resultfrom this workshop)

½ Second Priority

- Viability of development of a performance-based database

145

- Response of degraded structures (including finite-element analysis techniques, possiblyleading to an International Standard Problem)

½ Third Priority

- Instrumentation and monitoring3

- Repair methods- Criteria for condition assessment

• The CSNI should review again in two or three years the topic of ageing of concrete structures.

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3 : Based on presentation of the Task Group report to the CSNI, it was recommended by

representatives from Belgium and Spain that instrumentation and monitoring should be a higherpriority, possibly a first priority.

146

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Based on a review of national and international activities in the area of seismic structural behaviourof nuclear power plants, the task group focused its discussion around several broad areas, one ofwhich was the effects of ageing and degradation on the seismic performance of structures, systemsand components.

It was concluded that studies are needed to evaluate the effects of ageing and degradation. Theresults of such studies will be important for life extension and in evaluating maintenance practices.The specific issues of interest are :

• experimental evidence to identify localised failure modes and demonstrate margin,

• inspection guidance,

• how and when inspections are needed criteria for corrective actions.

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During this workshop a delegate of the Utility of each attending country4 gave a presentation of theprestress behaviour in their containments (or prestressed concrete reactor vessel). Some generalpapers on the creep and shrinkage of concrete were also presented.

The main elements resulting from the synthesis are :

• The reports from 12 countries have provided a large variety of cases including (see table):

− grouted system (72 containments) and un-grouted system (88 containments),

− Freyssinet, BBRV, VSL and SH type cables.

• The accumulated ageing experience was ranging from 4 to 39 years.

• Prestress loss > Predicted on some containments in 3 countries.

• Prestress loss < Predicted on some containments in 5 countries.

• No report on loss vs prediction from 4 countries.

• It appears that the prestressing loss is not a generic problem and it can not be associated with atype of prestressing cables or prestressing system. it is not dependent on the geographicallocation either.

• More information is required in each case to define the reason for a higher loss than anticipatedand its evolution in time. It is not clear yet if the loss is only quicker than anticipated or will alsoend up to be higher than anticipated at the end of the service life.

• Among the main reasons for higher than anticipated loss were: early application of prestressing,high temperature in the vicinity of the cables, inadequate codes provisions.

Three main topics were analysed:

• Corrosion protective media:

− Presently available products are reliable.

− Risk factor: no risk identified.

− Action suggested:

- gather from Users the brand names used with feed-back of performance,

- create a database and distribute to interested parties,

- develop guidelines.

4 : Belgium, Canada, China, France, Germany, India, Japan, Pakistan, Russia, Sapin,

Ukraine, UK, USA.

148

• Lift-off test:

− Mechanical Procedures (callipers, gauges insertion) are not available.

− Computerised reading instrumentation is reliable and available.

− Pressure cells should be abandoned.

− Risk factors: no risk identified with proper instrumentation.


- gather from Users specific data on instrumentation used and performance feed-back,

- create a database and distribute to participate.

• Loss of prestressing force:

− The loss of prestressing in containment is different from other concrete structures.

− The phenomena is local, not generic and can not be associated with a type of prestressing orcable system or geographical location.

− Additional and more specific data is required to understand each case and be able to drawgeneral conclusions.

− Risk factors:

- it appears not to be a plant life limiting factor since its affects the leak tightness only,which can be controlled,

- there are so far no available repair methods for grouted cables.


- gather and harmonise data available to make it interchangeable,

- identify and fill gaps,

- distribute data to the participants.

150

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Seismic re-evaluation, typically carried out at intervals of approximately ten years, applies inparticular to older plants which may not have been designed to a lower standard.

The status report is based on the responses to a questionnaire issued to all members of the SeismicSub Group.

It is to be noticed that seismic re-evaluation does not deal directly with ageing : any significantdegradation is expected to be identified and rectified by other safety processes.

Among the conclusions and recommendations drawn in this report those which are more relevant toolder plants are listed below :

• Several different re-evaluation methods are employed in practice, including PSA, margins anddeterministic analysis. A seismic walkdown is a key feature of re-evaluation in most countries,usually based on the SQUG (Seismic Qualification Utilities Group/US) criteria, or similar.

• Input motion levels, seismic categorisation, analysis methods and assessment criteria applied inre-evaluation are generally similar to those specified for new plant, ensuring that similarstandards of seismic resistance are available for plants of all ages. To understand plantbehaviour for severe accidents or risk estimates, more realistic criteria are used.

• All responding countries take account of the as-built situation, and the majority of countriesinclude in situ inspection as part of this process. Most countries rely on the original specificationfor material properties rather than in situ evaluation.

• Although nearly all countries are satisfied with the evaluations which have been completed todate, several aspects of the process are identified which might be improved in the future.

These include :

− Suitable hazard definition,

− Improvement in fragility determination,

− Improvement in equipment selection,

− Overall methods and criteria, including selection of structural models.

• The amount of modification carried out as a result of the re-evaluations can be extensive,including strengthening of buildings, walls, anchorages and equipment/pipework supports, andthe removal of seismic interactions.

• The physical modifications resulting from the re-evaluation have led to significantimprovements in the seismic ruggedness of older plant in most countries.

151

• This study suggests that there are several areas of the seismic re-evaluation process which couldbe considered in the future for the mutual benefit of the member countries. These include :

− The pursuit of a more detailed understanding of the benefits and disadvantages of thevarious methods of re-evaluation in particular circumstances.

− The definition of the scope of plant to be selected for the re-evaluation process, includingthe consideration of secondary hazards induced by the seismic event.

− Definition of the criteria for re-evaluation; including hazard and fragilities, and their relationto overall reliability.

− The role and scope of the peer review process, together with the qualification of experts inspecialists fields.

− The strengthening of plant.

− The incorporation of operational and research data/experience into the re-evaluation process.

− Identification of areas on new research that could provide benefits and improvements for there-evaluation process.

152

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The report focuses on the application of NDE to support the engineering assessment of the safety-related concrete structures found in nuclear power plants and nuclear chemical plants. Thesereinforced and prestressed concrete structures are characterised by thick sections, heavyreinforcement, liners and external cladding, and limited accessibility (often single sided accessonly).

Corrosion of steel reinforcement is recognised as being one of the commonest causes ofdeterioration of reinforced concrete structures, which has resulted in significant research anddevelopment effort on methods to detect corrosion. Initiatives in this area have generally involved acombination of visual and electro-chemical techniques. However, these techniques may not beapplicable to the thick-sectioned or inaccessible structures found in nuclear power plant, and it isthis particular aspect which is covered in the report.

NDE has potential applications in three key areas in the management of safety related concretestructures :

• determination of as-built structural details,

• detection of flaws,

• characterization and quantification of flaws (means for monitoring concrete ageing).

The report focuses on what was believed to be the more promising techniques (RADAR, acoustic,radiography) for assessing the condition of existing structures. Complementary assessment toolssuch as instrumentation/systems for continuous monitoring of structural performance ordestructive/semi-destructive tests were not considered in any detail.

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In recent studies radar technology proved to be the cheapest and easiest method for mappingreinforcements but neither characterisation of flaws by dimension and material nor crack detectioncould be demonstrated. Nevertheless radar has significant potential for development by way ofsoftware for signal and image processing to improve resolution around and immediately beyond thefirst reinforcement. This can be expected to achieve a capability to detect and locate furtherreinforcement (depending on rebar spacing), and to resolve gaps in reinforcement. It thus offersconsiderable potential in dealing with thick sections, where reinforcement is not too heavy. Radar isunlikely to approach radiography in terms of detailed inspection of individual reinforcing bars,including loss of section.

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Acoustic wave transmission can be used to obtain information about the physical condition ofconcrete structures. They are used either to characterise the properties of the concrete by wavespeed measurements or to locate and identify discrete objects and flaws in the concrete bytransmission and reflection of stress waves.

153

Acoustic testing methods or stress wave propagation methods encompass all forms of testing basedon transmission and reflection of stress waves : Ultrasonic Pulse Velocity (UPV), Ultrasonic PulseEcho, Surface Waves, Impact Echo, Acoustic tomography.

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Radiographic techniques are not widely used in inspection of safety related concrete structures innuclear plant because they are unsuited to penetration of the thick (> 1 m) sections commonlyencountered, require dual sided access and can present significant operational difficulties.Nevertheless, gamma radiography (together with gamma scintillation techniques) has been effectivefor determining internal damage in thin, lightly reinforced structures. It is of particular value fordetection and measurement of reinforcement/prestressing tendons and voids but can only be usedfor structures less than a metre or so in thickness. Newly developed high-energy x-ray acceleratorsare portable and compact. These allow practical inspection of concrete up to 1.2 m thickness.

Real time radiography is a possible area for development which could be combined withtomographic techniques to obtain improved results.

In its appendix 3 the report reviews a wide range of techniques applicable to inspection of concretestructures :

• mechanical : Schmidt Hammer, Falling Weight Deflectometer, Mechanical Gauges,

• electrical : Covermeter,

• electrochemical : Half-cells potential measurement, Resistivity,

• electromagnetic and nuclear : Radar, Radiography, X-Radiography, Gamma-Radiography andscintillation counting, Neutron radiography, Thermography,

• acoustic methods : Ultrasonics, Laser ultrasonics, acoustic emission, modal analysis,

• minor inspection and monitoring techniques : Holography, Strain gauging, Barkhausen noise,Automated Visual inspection, X-Ray tomography.

The main conclusions and recommendations are :

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NDE techniques have been used successfully on a variety of reinforced and post-tensioned concretestructures, notably highway and reservoir structures. However, there is limited experience of theiruse to evaluate typical nuclear safety related structures having thick sections, steel liners or accessto one side only.

There is a general lack of confidence in the techniques because there is very little independentadvice in their applicability, capability, accuracy and reliability. The information obtained bytechniques such as RADAR, ultrasonics, stress wave and radiography appears qualitative ratherthan quantitative and there is concern that NDE procedures lack the necessary qualification topermit their use on safety critical structures. There is no authoritative international guidance orstandard for NDE of concrete structures.

NDE of concrete structures is often based upon equipment developed for other materials andtechnologies, e.g. examination of steel, evaluation of ground conditions. Other industries aredeveloping equipment specifically for civil engineering applications and at the recent OECDworkshop a number of relevant national and European programmes were identified. The nuclearindustry maintain its awareness of developments and should seek to influence the development ofequipment.

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154

• More formal liaison with other industries that use NDE techniques should be established andopportunities to work with suppliers to influence the development of new equipment should besought.

• Experts should be identified to monitor national programmes with the aim of improving theunderstanding of the availability and capability of NDE techniques within the nuclear industry.

• CSNI should review this topic in approximately 3 years.

• At the time of the review, consideration should be given to quantification of the capabilities ofNDE techniques by means of a standard test specimen specification.

• As a longer term issue qualification should be considered.

155

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The objective of this report is to survey ageing problems, ageing management practices andresearch with regard to organic materials in NPPs and to identify areas where further work isrequired. Because of the complex nature of the subject the in depth survey is focused on electricinsulations, seals and lubricants under normal operation.

The structure of the reports is as follows :

• Section 2 covers the technical background of ageing of organic materials, including an overviewof the types of organics materials used in NPPs (see fig.), a discussion of safety aspects ofageing, a characterisation of stressors and main ageing mechanisms as well as a brief review ofinternational operating experience.

• Section 3 provides an overview of the range of ageing management methods which are currentlyin use. This includes qualification procedures, environmental monitoring, lifetime assessment,in-service inspections and maintenance.

• Section 4 covers information about national and international research programmes with regardto their objectives, the methodology for ageing testing and prediction, the compilation of dataand co-operation.

The OECD/IAEA Incident Reporting System (IRS) Database gives a broad estimate of the safetysignificance of ageing of organic materials in NPPs. For the majority of organic materials used therehas been little evidence of significant degradation. However some safety-related incidents were re-ported; in the order of their frequency, information was found about cases of ageing degradation of :• Seals• Lubrication• Cable insulation• Valve seats• Valve Diaphragm• Contactor• Expansion Joints

The degradation was primarily caused by thermal ageing.

The conclusions and recommendations of the report are :

&21&/86,216

The following general point have come out of this survey on the ageing of organic materials used innuclear power plant.

• Most research and international collaborative programmes in the area of ageing and degradationstudies have been on cable materials, and to a lesser extent, on sealing materials. Information ofthe ageing of other organic components is fairly limited and mainly arises from formalqualification procedures rather than any study of degradation mechanisms.

• Ageing behaviour is strongly dependent on the specific formulations used in the organiccomponents. In many commercially available organic components, the formulations will includea high proportion of additives and fillers which will affect the ageing behaviour of the baseorganic material.

156

• Realistic accelerated ageing is complicated by a number of factors which need to be understood,both for research work on ageing degradation and for qualification of components for use inNPPs. These factors include :

− dose rate effects (which are known to be significant in some polymeric materials used inseals and cables)

− synergy between thermal and radiation ageing (the ageing observed is not always simplyadditive)

− interaction between different organic materials within the same component (e.g. jacket andinsulation materials in cables are often of different materials). The degradation productsfrom one polymer can affect the ageing behaviour of the other.

• The development of practical condition monitoring methods for cables has made considerableprogress over the last few years and work is on-going. There is still a need for in-plant assess-ment of these methods and work on their use in assessing residual life. Very little is currentlyavailable for condition monitoring of organic components used in NPPs other than cables.

• Despite the large body of research carried out on organic materials, there are still no definitionsof end of life criteria that are realistic in terms of the functionality of a component.

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There are several recurrent themes on requirements for future work in the replies to thequestionnaire and in reports on research work on ageing of organics. There are :

• further development of condition monitoring methods, particularly those that are non destructiveand can be used in-situ in NPPs

• methods for assessment of residual life from condition monitoring data and accelerated ageing• comparison of lifetime prediction models with real-time ageing, using predictions based on

actual service conditions not design basis data• realistic failure criteria that relate to loss of functionality.

Application of organic materials in NPPs and important properties :

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Sheath materialsDielectric insulation,Strength/Flexibility

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Connectors,Terminations,Switches,Insulation in motors andtransformers,Terminal boards

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Leakage rate,Compressibility,Strength/Flexibility

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Viscosity,Friction coefficient

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Adhesive strength,Elasticity

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This document presents the proceeding of a Specialists Meeting hosted by IPSN under thesponsorship of WANO and CSNI PWG1 and PWG 3 jointly. The conclusions are as follows :

Thermal fatigue due to stratification and mixing of hot and cold water is a recurring phenomenon,but with a relatively low frequency. It can affect safety related piping of many different systems(ECCS, RHR, AFW, pressuriser surge line and safety/relief valve discharge lines) at variouslocation. It may happen that these phenomena are not included in the design conditions. Recently,similar events have occurred in sections of piping in the primary circuit and therefore the plantowners and the safety authorities are now alerted over some conditions not being consistent withlicensing basis and inspection commitments. Of particular concern are those problems arising fromunanticipated thermal fatigue in unisolable piping connected to the reactor coolant system. None ofthe mentioned incidents led to radioactivity propagation to the environment, but the safetysignificance of these results from the leakage of primary coolant through the second barrier butinside the reactor containment.

There is a need to develop further accepted methods to identify locations with potential risk ofthermal fatigue. There are proposals for simplified screening criteria based on semi-empiricalmodels to determine the areas where there is risk of thermal fatigue, but these are not generallyaccepted. As there are many uncertainties, it is possible to consider a probabilistic treatment,although the data for this are also limited. As first step in this direction, the use of best estimateanalysis methods should be considered, so that attention is focused on the areas most at risk forthermal fatigue, although these may be difficult to define. Then, at those locations, monitoring oftemperature and of pressure if necessary should be implemented. In addition, periodic verificationof the leaktightness of the nearest valves could further reduce the risk of thermal fatigue.

Monitoring of temperature fluctuations can be seen as an important part of the defence and, atpresent, it remains the most reliable method to avoid unanticipated incidents. At present, there aredifferent strategies in use and no single method provides defence. There is a need for combiningredesign and revised operating practices. A small internal cold water leakage into a hot section ofpipe can lead to a quick propagation of cracks by thermal fatigue in some sensitive zones.Manufacturing process probably has a very important impact on the rate of crack initiation. It is notpossible to draw up simple criteria for such parameters as allowable valve leak rates or limitatingpipe diameters, as there is a great variety in the systems and operating procedures.

Concerning the experience from non-destructive examinations, ultrasonic testing gave numerousfalse calls, and for certain geometries and material conditions, performance of present NDEmethods to detect fatigue cracks is limited. If a greater reliance is to be placed in NDE, thedevelopment of qualified methods is needed.

Thermal fatigue problems can be seen as challenging issues for plant owners. Because of the highpotential impact on safety, cost and radiation exposure, these issues have to be addressed moreeffectively. This is possible only with a very close co-operation between designer and plant owner,and among plant personnel, between maintenance and operation staff. Having all these peopleworking together is the key point to keep the risk of thermal fatigue under control and, moregenerally, to ensure safe and effective plant operation. There is a need to encourage co-operationbetween the different disciplines to solve the problem.

158

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Degradations refers to changes in the material with time and operation, and covers the concrete,steel and liner. Both functional and structural aspects are considered here. The conclusions andrecommendations are as follows :

&21&/86,216

• Finite Element methods are widely used to meet regulatory requirements for assurance ofstructural safety.

• For most analyses of concretes structures, simplified methods or linear elastic analyses areadequate but for realistic response predictions non-linear analyses are often needed, especiallyfor high temperature applications or prediction of local failures.

• The application of non-linear FE analysis to degraded concrete structures is considered to be arelatively new research subject. There is limited information available on non linear behaviourof concrete. A valid non-linear analysis depends on a constitutive model that can adequatelyrepresent the behaviour of concrete beyond its linear range, and appropriate materials data.

• Other industries are currently pursuing FE analysis of degraded structures.

• Some scatter and uncertainty in the results have been identified.

• Three dimensional FE calculations for reinforced concrete require a good understanding of thebehaviour of concrete structures, and experience to judge the validity of the results.

• Degraded structures have special features of material behaviour and structural modelling thatneed to be considered.

• With few exceptions, analysis of initially degraded concrete containment structures has not beenadequately investigated.

• There are very limited experimental data on degraded concrete structures to permit validation ofnon-linear FE analytical method.

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• Consideration should be given to future benchmarks/ISPs to improve the validation of themethods.

• Some improvement is needed in the analyses where thermal and mechanical processes must becoupled to permit specific applications to degraded structures.

• Test results for the shear transfer in cracked reinforced concrete panels would be useful todevelop/refine models for degraded structures.

• The use of probabilistic methods, including stochastic FE methods, should be pursued.

• The use of instrumentation on real structures to validate FE codes is a topic that it would beuseful to pursue in the proposed PWG3 workshop on instrumentation of concrete constitutivemodels.

159

• There is an need to do further work to better determine material property data and constitutivemodels.

• Consideration should be given to the development of indicators of local and global ageing.

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This document is based on the proceedings of the workshop on this topic held in 1997 (JointWANO/OECD-NEA Workshop : prestress loss in NPP containments, NEA/CSNI/R(97)9).

The conclusions and recommendations are as follows :

1. Present experience suggests that the current methods for the prediction of the loss of tendonprestress are generally satisfactory.

2. The nuclear industry has adopted regulatory and codified methods for predicting the loss ofprestress in nuclear power plant (NPP) prestressed concrete containments from international andnational standards that are not necessarily specific to nuclear design. The application of thedifferent methods to a specific case is likely to lead to significant differences in the predictedlosses.

3. Theoretical and experimental research have established the importance of understanding howchemical, hygral, mechanical and thermal factors influence the short term and long termbehaviour of prestressed concrete. In particular, they have differentiated between creep, dryingshrinkage and relaxation of prestressing steel and identified the interdependency of thesephenomena. However, research has, as yet, failed to formulate a universal and reliable model forpredicting both short and long term loss of prestress in actual prestressed concrete structures.Current and proposed activities aimed at improving the prediction of loss of prestress include :the creep behaviour of concrete in a biaxial or multiaxial stress field, standardisation of creepexperiments to provide reliable data ; experiments on the effects of temperature on prestressingsteels, and the development of approximate formulae and both empirical and semi-empiricalmodels to improve the prediction of shrinkage and creep in concrete, and relaxation of steel.

4. Improved and simplified simulations of creep and shrinkage phenomena that can account for theenvironment and loading history of prestressed concrete containments and pressure vessels willassist : the development of design regulations/standards; the choice of concrete mix ; thedevelopment of relevant monitoring programmes, and ageing management including plant lifeextension.

5. Prestressed concrete containments and pressure vessels use both grouted (bonded) tendons andungrouted (unbouded) tendons. The workshop considered the relative merits of both systems.

− Grouted tendons. The cementitious grout : surrounds the tendon in an alkaline environmentthat will inhibit corrosion of the steel, and prevents the ingress and circulation of corrosionfluids. In case of break of a tendon, due to the bond with the grout, part of the prestressremains transmitted to the concrete.Therefore grouted tendons are less vulnerable thanungrouted tendons to local damage. They reduce the risk of the containment being by-passedvia tendon ducts, particularly important where the containment is unlined. However, groutedtendons can not be visually inspected, mechanically tested or re-tensioned in the event ofgreater than expected loss of prestress.

− Ungrouted Tendons. Prestressing force is transmitted to the concrete, primarily, at thelocation of the anchorages. Corrosion is prevented by organic, primarily, at the location ofthe anchorages. Corrosion is prevented by organic petroleum based greases or corrosioninhibiting compounds. These are either applied to the surface of the tendon prior toinstallation or injected into the tendon duct following completion of the stressing sequence.

161

Some countries use a combination of both coating and injection. Tendons can be removedfor visual inspection/replacement ; mechanically tested in-situ ; and retensioned to maintainprestress. Ungrouted tendons are more vulnerable than grouted tendons to local failure andcorrosive fluids can circulate along the ducts. Ducts may provide a route for containment by-pass in unlined containments, although the practice of keeping ducts filled with corrosionprotection medium reduces the likelihood of by-pass.

6. Experience presented at the workshop indicates that comprehensive and regular monitoring ofthe behaviour of containments and pressure vessels at operational plant assist our understandingof the cause of loss of prestress. Containments around the world include instrumentations tomeasure : anchorage loads ; concrete strain ; structural geometry ; concrete temperature ; andsurface cracking. Data collected from more than 150 structures aged between 3 and 40 yearsindicate that, for the majority, loss of prestress has been less than predicted. However, for somecontainments, losses have exceeded predictions. Measured losses vary from containment designto containment design but significant differences have also been observed betweencontainments in the same design series. The variation in measured losses has been attributed to anumber of factors including : concrete composition ; aggregate type ; the presence of a liner ;high relaxation of steel tendons ; concrete temperatures ; loading history and the environment.Regulatory and codified prediction techniques do not necessarily account for such factors.

7. Many plant include direct measurement of tendon loads at the anchorage. A number of papersreported problems with the reliability and accuracy of tendon load measurement. The use oftendon load to interpret loss of prestress requires careful consideration of the method used tomeasure the load and the design of the prestressing system.

8. Nuclear containments and pressure vessels are designed with large margins on structuralintegrity. Therefore, a higher than expected loss of prestress does not necessarily jeopardise theintegrity of the structure. However, under accident conditions the margin on precompression ofthe concrete is reduced and therefore there is an increased risk of cracking. This may result in acorresponding increase in the leak rate of unlined containments. Periodic testing of thecontainment is used to evaluate its leaktightness.

9. The workshop discussed the corrosion protection media used for containments and pressurevessels having ungrouted tendons. For systems where the tendon duct is filled with protectionmedia, greases have been developed that optimise : viscosity, resistance to penetrating concrete; water displacement ; alkalinity and electrical conductivity. For systems using coated tendons,with time the grease looses its lighter oil component but the residue is still capable of providingcorrosion protection to the tendons.

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1. Information on direct load measurements (in particular their accuracy and reliability) should beconsidered at the future PWG 3 concrete group instrumentation workshop.

2. Current research activities and operational experience should be followed to ascertain whetherpresent uncertainties in predicting losses can be reduce

162

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This question is still open. The conclusions and proposals of the document are as follows :

• Ageing is permanently present and its effect is globally to decrease the seismic resistance butthe quantification is not easy.

• Structures seem to be less sensitive than equipment.

• It is believed that widespread ageing is necessary before there was an effect. Definition of athreshold could be interesting.

• Distinction should be made between normal and abnormal ageing ; the latter must be solved asan ageing question.

• Some parts, such as supports and anchorage, are loaded only by seismic loads and are moresensitive to this question.

• Maintenance and walkdowns decrease the effects of ageing. Plans must be adapted to seismicconditions. Inaccessible parts may be a problem.

In order to continue this topic, possible actions may be the following :

• Write a list of ageing sensitive structures and equipment, starting from existing PWG3 or otherpublications and give judgement about their seismic potential risk. Potential failure modes willbe indicated. This list will drive future work on the subject.

• Irradiation embrittlement of internals must be examined.

• The assessment of inaccessible parts must be studied (EPRI is working on the subject).

• Maintenance programs should examined from a seismic point of view.

• Inspection techniques for anchorage and sensitive parts must be improved.

163

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This document takles the technical aspects of the basis for a long-term operation.

It focuses on the safety related barriers and the internals of the pressure boundary. The managementof ageing of active components is understood above all as a maintenance issue.

It gives the following conclusions and recommendations :

The safe and reliable operation of light water reactors has been proven for almost 30 years. It hasbeen demonstrated that even for components with high field of radiation the access for repairmethods is satisfactory. Although some serious damage mechanisms have limited the lifetime ofsome components, replacement could be performed without excessive burden to the availability ofthe units.

In view of the general technical experience the factor of extrapolation with respect to time or usageare a factor of two or less if one considers operating times between 40 and 60 years.

Due to the drastic changes in industry influenced by the declining demand for new nuclear powerplants it is to be expected that in most countries the industrial infrastructure is being sized down tomaintain the requested level of service only. In many technical areas continuous development willtake place driven by other industrial development than nuclear. This is certainly to be expected forinstrumentation and control but also in the areas of civil engineering, material production andwelding technology along with the technologies for surveillance, testing and inspection.

To provide a technical basis for long-term operation of nuclear power plants it is necessary to :

• update the individual plant documentation to avoid gaps in knowledge caused by thereorientation of industry,

• enforce clubs of users of similar technology internationally,

• establish a system of information retrieval to bridge gaps between today’s and previous designand manufacturing standards,

• increase the flexibility of the quality assurance system to qualify products manufactured to otherstandards for plant specific use.

164

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In response to interest shown by the Nuclear Energy Agency (NEA), Principal Working Group 1(PWG-1) of the Committee on the Safety of Nuclear Installations (CSNI) conducted a generic studyon the effects of ageing of active components in nuclear power plants. Representatives from France,Sweden, Finland, Japan, the United States and the United Kingdom participated in the study bysubmitting reports documenting ageing studies performed in their countries. This report consists ofsummaries of those reports, along with a comparison of the various statistical analysis methods usedin the studies. The studies indicate that with some exceptions, active components generally do notpresent a significant ageing problem in nuclear power plants. Design criteria and effectivepreventative maintenance programs, including timely replacement of components, are effective inmitigating potential ageing problems. However, ageing studies (such as qualitative and statisticalanalyses of failure modes and maintenance data) are an important part of efforts to identify andsolve potential ageing problems. Solving these problems typically includes such strategies asreplacing suspect components with improved components, and implementing improvedmaintenance programs.

The conclusions are reproduced below.

&21&/86,216

All participating countries have reported ongoing activities in their countries related to investigatingnuclear power plant ageing, indicating they recognise the importance of understanding andmanaging ageing. The Japanese report, however, makes clear their position that, for their plants,there are few events attributable to ageing, because deteriorated or malfunctioning components orparts are replaced when they are discovered during periodic inspection or are replaced as a result ofpreventative maintenance, often in response to problems experienced at other plants.

The information provided by the participants indicates that for most cases, the overall maintenanceand surveillance programs serve to produce relatively constant failure rates for the activecomponents. In addition, information provided on the evaluation of plant performance and safetysystem reliability confirm that ageing of active components is being adequately managed at theplant and system level, in that degradation of plant performance or reduced reliability have not beenobserved.

In a few cases, the failure rates increased with age, indicating a time-dependent contributor to thefailure parameter. This is true particularly of the case of the 6.6-kV circuit breakers for the FrenchBugey plants, where the increase was significant. The problem was recognised, and then correctedby modifying the ageing part and implementing an improved maintenance program. In the othercases, the indications of an increasing trend of the failure was in the incipient stage and required

165

qualitative and statistical analyses to infer the trend, or the increases were not great enough toproduce a noticeable effect on the performance of the system.

Several of the participants reported developments of statistical methods for evaluating failure datato provide early detection of increasing failure rates. The countries use similar statisticalapproaches. Use of estimates from binned data is natural. For modeling trends of failures in time,the nonhomogeneous Poisson process has been used most heavily, although some analysts wish tomake the model more realistic. Finland developed analytic methods for qualitative analysis offailure modes at the item level and used statistical methods to identify the recurrence of failuremodes.

The various statistical methods were shown to have the potential for monitoring the effectiveness ofmaintenance and surveillance programs; however, use of the methods does require extra attention tothe mathematical detail. The researchers in one case reported that trend analysis tools need to becapable of considering the times between failures instead of the number of failures of different agegroups. They also reported that an important step for the evaluation of ageing is to establish routinesfor merging and grouping age-related data sources and failure modes for later meaningful statisticaltreatment.

Several reports indicated that accurate calculations of equipment failure rates are difficult becauseof problems inherent in failure data. The problems identified include:

• Difficulty in establishing the population or the total number of demands that serves as the basefrom which to calculate the failure rates. (Although the failures are reported, the number ofpieces of equipment or the number of demands without failures are mostly not reported).

• Difficulty in establishing the time in service or the number of demands at failure. (Often themaintenance records are for an equipment function or location and do not provide data on timeof installation or replacement).

• Difficulty choosing between obtaining a set of data that is fairly homogeneous but representsonly a few failures, and obtaining a larger set of data that is less dominated by randomvariability but is more difficult to interpret because of differences in hardware characteristics,operating conditions, failure modes, etc.

• Difficulty in determining accurately the instances when preventative maintenance preventsfailures from occurring.

• Use of periodic tests for determining the failure rate when the conditions for the safety functionare much more severe than for the test. (Particularly for valves, the periodic tests may not be agood indication of their ability to open or close as required during an accident condition withhigher differential pressure and temperature).

Several of the reports identified lack of component-specific data from installation to removal as themajor obstacle to more specific ageing and maintenance studies. Ageing studies and maintenanceprograms could be more effective if failure reports included:

• the cause of the failure

• the date the equipment entered service or was replaced

• modifications, preventative maintenance, and corrective maintenance

• identification of the component that failed

• the function of the equipment degraded by the failure of the component.

One report, HPCI System Performance, compared system unreliabilities observed from experiencewith the corresponding unreliability estimates from the plant probabilistic risk assessment (PRAs).For a single injection without recovery, the observed unreliabilities for about half the plants were inreasonable agreement with those determined from the PRAs, but for the others the observed

166

unreliabilities were greater than three times higher than those from the PRAs. If recovery isincluded, all observed unreliabilities are in good agreement with those from the PRAs.

The report for the ageing assessment of circuit breakers and relays reported that the observedoverall failure rate for relays in the French plants was 5.2 x 10-9 h-1, whereas the value adopted forthe IPSN probabilistic safety studies was 3 x 10-7 h-1. However, the report cautions that becausethe consequences of failures of relays is often not large, it is possible that not all the failures werereported.

Evaluation of the failure data indicated that in some cases the current testing and inspection may notbe focusing on the dominant contribution to unreliability of the equipment. Also, in some casesthere are recurring problems with reliability, in particular with steam-turbine-driven pumps andmotor-operated valves, where additional effort for correction is warranted.

The study indicates that detailed qualitative and statistical analyses of failure modes andmaintenance dates can identify potential ageing problems in active components, and can provide abasis for preventive actions to manage those problems when adequate failure data are available andwhen the data are used for systematic analysis.

167

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This report was prepared by the Task Group on Computer-based Systems Important to Safety of thePrincipal Working Group No. 1. Canada had a leading role in this study.

Operating and Maintenance Experience with Computer-based Systems in nuclear power plants isessential for improving and upgrading against potential failures.

The report summarises the observations and findings related to the use of digital technology innuclear power plants. It also makes recommendations for future activities in Member Countries.

The conclusions are reproduced below.

&21&/86,216

1. Feedback of operating and maintenance experience is recognised as an important input tofailure analysis associated with complex systems such as computer-based systems. The processof feedback would provide designers with information on systems failures, unforeseenscenarios, or unanalysed configurations. Following plant start-up, the use of the operatingexperience have led to reconfiguration of system components. Several views on ways in whichthis feedback can be achieved have been presented.

2. Safety Assessments by regulators use the operating experience in an assessing computer-based safety performance prior to installation in areas such as electromagnetic interference,software reliability or human-machine interface. Assessments were also extended tomanagement of modifications after plant start-up.

3. Programmable Logic Controllers (PLCs) appear to offer good potential for successfulcharacterisation of software performance due to their widespread use in safety and non-safetyrelated applications within and outside the nuclear industry. They can provide a sizeablepopulation for observation.

4. While computer-based system failures cause few significant safety events, they could causecommon cause failures leading up to significant events.

5. Software modification during operation is one of the major sources for software errors.Maintenance experience indicates that adequate documentation is essential in reliablyperforming changes in the software.

6. Preventative maintenance concept extends to computer-based systems in different ways,such as the inclusion of self diagnostic capabilities or saving memory data on restart to analysethe cause of an initial stall.

7. Computer equipment situated in a control room are more likely to be affected byenvironmental stressors than by design basis events. Primary initiating events could be loss ofheating, ventilating, and air conditioning, water spills, or use of fire suppression water. Whilerising temperature, for an analogue system, causes a loss of calibration accuracy, it can causemore serious effects on digital equipment including failure to perform their function at all. Moresevere errors were found to be caused by EMI/RFI.

8. A significant fraction of all errors resulting from the application of environmental stressorsis communication errors. Many of these errors were time-out errors or corrupted transmissions,

168

indicating failure of a computer to receive data from an associated multiplexer, optical seriallink, or network node.

9. Ageing does not appear to pose a significant design concern for digital systems because theequipment is installed in a mild environment and because it is accessible for monitoring,calibration, and replacement. Consequently, the equipment can be expected to be serviced orreplaced as necessary throughout the plant life. The installed equipment can thus be assumed tohave like-new performance.Seismic qualification of digital components does not appear to pose any unique qualificationissues. Surface-mounted integrated components are recognised as rugged components and areroutinely used in applications such as automobiles, aircraft, and portable electronic equipment inwhich accelerations typically exceed that of a design basis earthquake.

10. Software errors constitute a significant number of failures of software. They may includecoding error, design error or V & V error.

11. Maintainability of computer-based systems depends largely on the quality factors oftraceability, completeness, consistency, simplicity, modularity and testability. The maintainerneeds to be able to fully understand the software before it can be changed. Improving upon thedeficient quality factors is required in order to improve maintainability.

12. Regulatory assessments of digital upgrades focus on issues related to reliability of softwareelectromagnetic capabilities (EMC) and training for human-machine interface.

169

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The report is based on the following considerations :

1. The aim of this paper is to set out the NEA’s view of why this subject is an appropriate oneto be handled under the NDC and to provide an overview of preceding work on it.

2. In 1987 there was a meeting of the NEA, held in co-operation with the IAEA, when it wasfairly comprehensively demonstrated, if demonstration were needed, that extension of plant lifeought to be a cheaper option than building new plants. At a time when competition in electricitysupply is becoming more severe, and the capital cost of new nuclear plants is so much higherthan, say, combined cycle gas turbine plants, the availability of the life extension option looksparticularly desirable. The current difficulties of finding new sites for nuclear power plants, andin some countries for almost any type of new power plant, only add to the desirability.

3. The preservation of nuclear capacity derives added attraction in the light of the wish to avoiddisturbance to economic growth that can be generated by over-reliance on outside suppliers, aswas the case in the 1970s, and of newer concerns to avoid increased dependence on powersources emitting carbon dioxide and other atmospheric pollutants.

4. All these considerations feed back into governments’ policies on medium and long termenergy supplies. Consequently governments are also concerned about the reality of the lifeextension option. There can be challenges to its reality from technical, regulatory and economicfactors and, perhaps more importantly, from the interplay of these factors. Other parts of theNEA as well as the IAEA provide a forum for exploring what, if any, are the technical factorsthat will limit component and plant life. They also provide fora for discussion of the proceduresfor licensing older plants in the context of updated regulations. The Symposium already referredto was the start, as far as we know, of international consideration of the economic aspects.

5. It appears clearly that one fourth of the reactors in the NEA Member countries, more than100, are over 20 years old and more than 40 reactors are over 25 years old.

6. The current status of the envisaged nuclear power plant life and some model calculation withdifferent assumption are given in the report. Plant life extension and its implications can befound quite clearly. The retirement of these plants over the next few decades could posesignificant challenges to the meeting of government policies on energy supply. Conversely,continued well-managed operation of existing nuclear power plants would be an enormous steptowards fulfilling the goals of energy security with environmental protection. If we assumefurther development of nuclear power plants in addition to the plant life extension, the total sumof world nuclear power installation would be maintained for a considerable duration.

7. In the period immediately after the NEA’s first efforts in this field it became evident that,because of the limited numbers of actual plant life extension at the NPPs, the concept of plantlife "management" was not well-developed. It also seemed that the reality of a relative cheaplife extension was not always obvious. In some cases the cost of gaining sufficient confidencethat regulatory requirements for continued operation could be met was seen as more than theexpected return on investing in life extension. Thus it was not cost alone that mattered but theprobabilities of the risks still remaining were very important. More information could reduce theuncertainties but acquiring information could itself be costly or might take such a time to

170

acquire as to prevent application for licence renewal being made before the plant had to go outof service.

8. During the ten year since then, we have seen a number of plant life managementprogrammes in Member countries, including major refurbishment works on the plants. Theinitial set of PLIM issues was closely related to the regulatory license renewal at the age ofthirty years old, however, the PLIM concept has been developed with wider perspective, i.e.total life management of the NPPs, not only license renewal, has been identified and realised innumber of countries. The development of thinking about these concepts is one of the reasonsbehind the setting up of the meeting.

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9. The main scope and objectives of the PLIM Expert Group were to assist decision makers atthe Member states’ government level in evaluating the economics and politics of plant lifeextension by providing a published report in which the following types of key issues areaddressed in broad terms.

− Rationale for plant life extension in different NEA Member countries (economics, financial,reliability, availability, safety, etc.);

− Financial and economic concerns including intermediate results at incremental time periodsof extension (5, 10, 15 and 20 years);

− Major technical concerns and constraints;

− Institutional issues in and public acceptance on the plant life extension;

− Need for and feasibility of creating data base which will encompass a wide range of data forplant life extension;

− Precedents in other NEA Member countries and IAEA Member states.

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10. From 1993 onward, although a considerable gap in 1994-5, separate work was carried out onan International Terminology for PLIM in five languages, based on the US-EPRI’s workpublished in 1993, with the co-operation of IAEA and CEC, and this work will be publishedlater this year.

11. The report consists of two different sections; i.e. a model approach, and national programmesummaries. The former includes various model elements of PLIM programme for owners;critical items and components of NPP in different types of reactors; and safety assuranceprogramme considerations including license renewal and periodic safety assessment issues. Thelatter consists of country report of their PLIM related activities. The report was aimed to satisfythe goals for phase I of the activity. It was envisaged that it would be followed by a phase IIwhich was expected to analyse the current requirements and the feasibility of obtaininginformation necessary for plant ageing and life management decisions, and also to specify theareas where governments’ support and international co-operation are desirable.

12. NEA Member countries have acquired number of PLIM experiences for different types ofreactors with different histories. Ad-hoc sets of information are being stockpiled piecemeal indifferent countries and different companies. Some of this has appeared at specialist conferencesand in the technical press. A wide range of information, has been presented together with moreor less sophisticated ways of analysing plant life decisions. Nevertheless, it cannot be safely saidthat there is an established set of general rules or guidelines. Rather it still seems that there is alack of integrated information for the future decision makers.

171

13. In view of the general energy policy concerns expressed at the beginning of this paper, it canbe argued that governments still have an interest in helping the development of efficientprocedures for plant life management. Exchange of experience, transfers of lessons betweendifferent countries could help to reduce uncertainties, point to the scale of effort needed andwhat sorts of effort are most useful, and generally add to the solidity of understanding of thefuture course of nuclear energy.

14. The presentations from experts from the following countries have been reviewed:

− Germany− Spain− the Netherlands− France− United Kingdom− Czech Republic− Japan− Belgium− Hungary− Canada

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1. A case was presented at the beginning of the meeting for the continued involvement of theNDC in the subject of Plant Life Management. It is a case that will not necessarily appeal to allMember countries. At a time of increasing demands on a tightening budget NEA do have toconsider whether there is really a role for the NDC, or whether the sort of studies commissionedearlier on should be done in some other, possibly more industrially-oriented, forum. Somecountries have developed lists of critical components and equipment of the NPPs though theymay not be disseminated internationally.

2. As a minimum it is suggested that the NDC should sponsor at regular intervals, say everysecond year, a workshop such as this one to gather and consider recent national and utilityexperience of plant life management activities, seen in their economic context. It is assumed thatactivities of the NEA and IAEA on the technical and regulatory aspects would continue anyway.It would always be open to such meetings to develop ideas for focused activities such as theproduction for consensus reports.

3. With the evolution of experience and practice, it is imagined that there might well be a needfor restructuring of the report but always in line with the original objectives of the activity. Asmall core editing group is desirable to complete the tasks involved.

4. The most ambitious objective of all would be to try to develop a set of guidelines but themeans and benefits of doing so seem difficult to judge as of today.

172

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Given the present restructuring trend in the electric utility industry, the issue of maintaining thenuclear power plants in operation while competing effectively with other forms of electricitygeneration is becoming increasingly important. While leaving aside other considerations, such asglobal climate change or security of supply, on a purely cost basis, it is of the utmost importance forelectric utilities to be able to prove the competitiveness of nuclear power. From this perspective, thesharing of knowledge on refurbishment costs which have been incurred or that are expected in thefuture is considered to be very valuable to countries with nuclear programmes. In order to achievethis objective, in 1997, a group of experts was set up within the OECD/NEA, inviting Membercountries to participate.

The study performed is a component of the OECD/NEA programme on Plant Life Management(PLIM) and the objective was to collect and evaluate nuclear power plant refurbishment costs dataand experience accumulated over the last years in participating countries. This information may beuseful to reactor operators faced with nuclear plant life cycle evaluations.

This report presents refurbishment cost data derived from experience and plans to implement PLIMprogrammes in ten OECD countries (Belgium, Canada, Czech Republic, Finland, France, Hungary,Mexico, the Netherlands, Spain, and the United Kingdom).

The information was collected by issuing a questionnaire prepared by the Expert Group and sent toMember countries in September 1997. The purpose of this questionnaire was to try and reflect theviews and concerns of all members of the Expert Group. The data provided by most countries wasfrom 1990 onwards and previous data was only given if country experts considered it to be relevantfor the purpose of the study. Refurbishment activities were reported for 89 reactors (69 per centPWRs, including VVERs, and the remaining 31 per cent shared by BWR, CANDU, Magnox andAGR types).

Most countries attempted to provide one or more specific motives for the implementation ofrefurbishment activities. Motivations were reported almost evenly among the four major categories(regulation, upgrading, economics and life management/extension). Some countries addedsummaries of existing regulatory policies for plant life management.

In relation to the scope of refurbishment, many different types of refurbishment activities werereported under equipment and systems. Under equipment, most of the activities are related to majorworks on sensitive equipment such as turbines, condensers and steam generators. Otherreplacements of ageing and inadequate equipment were also reported. Under systems, countriesreported considerable improvements in fire prevention and cooling systems.

Several countries provided very relevant information on costs of refurbishment activities whichenabled the formulation of cost ranges for major items replaced or expected to be replaced as part ofthe life management programmes. However, not enough information has been provided to estimatemanpower, financial resources and time for implementing regulatory policies on life extension.

173

The analysis of the reported cost data is limited due to:

• the diversity of approaches towards refurbishment activities, related regulatory processes andlife management among reporting countries;

• the difficulty in classifying data according to motivation;

• the different definitions used for refurbishment costs and its components;

• the different cost accounting methods used by nuclear plant operators;

• the difficulty in assigning the cost to a particular unit or a series of similar units; and

• the different periods in plant life for which data has been reported.

Given these constraints, the cost data are presented in summarised form on a country-by-countrybasis and they are analysed on a per unit, per net capacity (MWe) and per component basis.Whenever possible, the cost data are disaggregated into periodic and non-periodic costs. These twotypes of costs are considered to be the components of the overall reported cost.

The costs are expressed in terms of 1997 US$ per net MWe capacity. The costs are presented byreporting country; by reactor type and by reactor age. 7KH�FRVWV�UHSUHVHQW�RQO\�WKH�UHSRUWHG�FRVWV�QRW�WKH�RYHUDOO�WRWDO�FRVWV��DQG�WKH\�DSSO\�RQO\�WR�WKH�VSHFLILF�WLPH�SHULRG�FRYHUHG�LQ�WKHVWXG\��). As the range illustrates the order of magnitude of the reported cost data,comparisons among countries or reactor types are not necessarily meaningful and should beavoided. The large ranges in costs illustrate the diversity of factors, criteria and approaches relatedto refurbishment activities and consequently the difficulty in performing detailed quantitativeanalysis.

The study confirms diversity of criteria, procedures and regulations with respect to refurbishmentactivities and life management or life extension programmes among reporting OECD countries.Although some qualitative trends can be identified on the basis of collected data, quantitativeconclusions are difficult to derive. In most cases, expenses related to refurbishment activities do notrepresent major cost components that could jeopardise the expected economic life of a nuclearpower plant. In general, reporting countries have chosen to implement refurbishment activities inorder to satisfy regulations, expand capacity, improve performance and safety, and extend theeconomic viability of nuclear power plants.

Further studies should consider the analysis of the effects of deregulation and privatisation of theelectricity sector at national and international levels in future plans for refurbishment activities andlife management/extension. In addition, studies should consider the role that the debate on climatechange could play in future refurbishment activities given the importance currently attached to thisenvironmental factor.

The conclusions and recommendations are reproduced below.

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Based upon the information collected and despite the fact that data from some OECD countries withmajor nuclear programmes have not been considered, the following conclusions may be derived:

• All reporting countries are implementing life management or life extension programmes whichinclude refurbishment activities performed in the past, currently being undertaken or planned inthe future.

• The study confirms diversity of criteria, procedures and regulations with respect torefurbishment activities and life management/life extension programmes among reportingcountries.

174

• Although some qualitative trends can be identified on the basis of collected data, quantitativeconclusions are difficult to derive given the diversity in definitions and approaches torefurbishment and the limited amount of reported data.

• Reporting countries have chosen to implement refurbishment activities in order to satisfyregulations, expand capacity, improve performance and safety, and extend the economicviability of nuclear power plants. These motivations are generally multiple or not mutuallyexclusive and therefore are not easy to classify or correlate.

• In general, nuclear units designed before 1975 and characterised by limited physical separationhave incurred relatively high refurbishment expenses.

• Refurbishment activities due to obsolescence are mainly in the areas of instrumentation,computer and control systems, and to a minor extent in spare parts.

• In most cases, expenses related to refurbishment activities do not represent major costcomponents that could jeopardise the expected economic life of a nuclear power plant.

• Even when refurbishment activities represent in absolute terms considerable investments, interms of $/kWe, this investment is significantly lower than investments for alternative electricitysupply options.

• Although refurbishment activities have been successfully implemented in all reporting countriesensuring different types of benefits, it should be noted that political reasons can determine earlyretirement of nuclear power plants even if they are operated safely and economically. Arepresentative example is the case of the Borssele nuclear power plant in the Netherlands wherelarge expenditures have been incurred for upgrading according to the latest standards butnevertheless the lifetime of the plant is currently limited by a political decision.

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Although this study has enabled the collection and evaluation of very valuable data, further studiesmay be necessary to permit a more comprehensive understanding of the evolution and trends inrefurbishment activities and related costs. Such further studies should consider:

• More precise definitions of the type and scope of data relevant to the analysis of refurbishmentcosts.

• The effects of deregulation and privatisation of the electricity sector at national and internationallevels on planned refurbishment activities and life extension programmes.

• Analysis of the evolution of refurbishment costs incurred during the entire lifetime of eachnuclear reactor.

• The role that the debate on global warning and climate change could play in futurerefurbishment activities given the importance currently given to these environmental factors.

175

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This document was aimed at giving a general knowledge on ageing to a wide technical andmanagerial audience.

The contents is as follows :

1. Introduction

2. Definition of ageing

3. Ageing and its relationship to the safety of nuclear power plants

4. Equipment qualification : defence against common cause failures

5. Methods of determining safety significant components and systems susceptible to ageingdegradation

6. Material ageing mechanisms

7. Detection and mitigation of ageing effects

8. Managing the impact of ageing on nuclear power plant safety - Examples of activities inmember states.

As relates to the safety section 3 very briefly describes the ways in which the ageing degradationmay affect the integrity of the defence in depth : this is mainly due to the potential increase in theprobability of component failures and to a higher probability of common cause failures. This isaddressed by the equipment qualification (section 4).

The practical methods outlined for investigating ageing problems (section 5) are :

• analysis of operating experience

• expert opinion

• probabilistic techniques for prioritization and for determining risk significance of ageing.

Section 6 presents in some extent the basic mechanisms of ageing degradation of metals, concretestructures and non-metals (plastics, elastomers, lubricants, ...) but the major NPP components arenot addressed.

Section 7 gives the general principles of the methods for timely detection and mitigation of ageingeffects (preventive maintenance, predictive maintenance, scheduled maintenance, reliability-centered maintenance).

While a little outdated this document still constitutes an introduction to the ageing concern.

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This document has two objectives :

• to provide advice on data needs and on an effective and practical systems for data collection andrecord keeping in connection with the ageing management,

• to facilitate the sharing and exchange of data on.

The data analysis and evaluation is outside its scope. The general data needs are divided into threecategories (see table) :

• Baseline information, consisting of design data and conditions at the beginning of the servicelife of a component ;

• Operating history data, covering system and component level service conditions and componentavailability testing and failure data ;

• Maintenance history data, including data on the monitoring and maintenance of componentsconditions.

Examples of selection of component specific data need are given for a Reactor Pressure Vessel, anEmergency Diesel Generator and Electric Cables.

Examples of effective data collection and record keeping systems and guidance for theimplementation of an advanced data collection and record keeping system are given in appendices.

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This technical report presents methodologies both for selecting plant components important tosafety whose ageing should be assessed and for performing ageing management studies. Thismethodologies are based on current practices of Member States leading in this field.

The table of contents is as follows :

1. Introduction

1.1. Background

1.2. Purpose of the report

1.3. Technical Safety Issues

2. Approach to the management of NPP ageing

2.1. Grouping of NPP components

2.2. General approach to the management of NPP ageing

2.3. Equipment qualification : an example of an ageing management programme

3. Selection of NPP components important to safety for ageing management studies

3.1. Selection process

3.2. Approach to prioritization

4. Methodology for ageing management studies

4.1. Phase I : Interim ageing study

4.1.1. Review of existing information relating to the understanding of component ageing

4.1.2. Documentation of current understanding of components ageing

4.1.3. Review of current methods for monitoring and mitigation of component ageing

4.1.4. Phase I - report : Interim ageing assessment and recommendations for follow-up work

4.2. Phase II : Comprehensive ageing study

4.2.1 Studies on understanding ageing

4.2.2. Studies on monitoring of ageing

4.2.3. Studies on mitigation of ageing

4.2.4. Phase II report : comprehensive ageing assessment and recommended application of results

5. Recommendations for ageing management pilot studies

179

5.1. Topical areas proposed for pilot studies

5.2. Selection of NPP components for pilot studies

5.3. Technical issues relating to the pilot studies

APPENDIX I Examples of ageing related component degradation and failure

APPENDIX II Ageing degradation mechanisms and susceptible materials and components

APPENDIX III Examples of summary results of ageing management studies from the UNNRC’s nuclear plant ageing research programme

APPENDIX IV Examples of condition indicator trending as a basis for mitigating componentageing

The ageing management process consists of the three basic steps :

1. selection of safety important plant components for which ageing should be evaluated ;

2. understanding dominant ageing mechanisms in the selected components and identifying ordeveloping effective and practical methods for monitoring and mitigating ageing of components(ageing management studies);

3. managing the ageing degradation in the selected components by effective practices andinitiatives in surveillance, maintenance and operations (proper design, manufacturing, storageand installation are also significant in the management of ageing).

This report focuses on documenting the methodologies used in Step 1 (selection of NPPcomponents) and Step 2 (ageing management studies). It also provides recommendations for ageingmanagement pilot studies of specific components.

The technical safety issues addressed in the document are presented in the following :

1. Which NPP components are susceptible to ageing degradation that could adversely affectplant safety ? Which of these components are renewable (by maintenance, refurbishment orreplacement) ?

2. What are the degradation processes of materials and components that could, if unchecked(i.e. if components are improperly maintained and/or not replaced), affect plant safety ?

3. Are current methods for testing, inspection, surveillance, maintenance and replacementadequate to detect and mitigate ageing problems before they significantly affect safety ? If not,what additional measures are needed ?

4. Are current analytical models and criteria adequate to evaluate the residual life of keycomponents and structures ? If not, what additional criteria and supporting evidence (data,analyses, inspections) are needed ?

5. How should structures and components be selected for comprehensive assessments ofageing and evaluations of residual life ?

6. What kinds of records and other documentation are needed to support effective ageingmanagement ?

180

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This Safety Report supplements the NUSS Code on Operation (No 50-0, Rev. 1, IAEA, Vienna(1988)) and the associated Safety Guides ("Management of Nuclear Power plant for Safe Operation:A Safety Guide" Safety Series No 50-SG-09, IAEA, Vienna (1984) ; "Periodic Safety ofOperational Nuclear Power Plants : A Safety Guide", Safety Series No 50-SG-02, IAEA, Vienna(1994)).

It provides information on effective practices relating to the implementation and review of anAgeing Management Programme (AMP) ; such a programme constitutes a systematic "umbrella" tocoordinate the programmes relating to the maintenance, in-service inspection and surveillance aswell as operations, technical support and external programmes such as R&D.

It deals with the organizational and managerial aspects ; technical aspects and economicconsiderations are not addressed. It emphasizes the systematic ageing management of long livedpassive components and structures that are not routinely inspected, maintained of replaced.

Three elements are taken in account : material ageing, technological obsolescence and humanaspects.

An AMP model is proposed, based on the experience of several utilities (see fig.). The report detailsthe responsibility of AMP participants, the implementation of an AMP and provides guidance forreviews of its effectiveness. An important contribution of this document is the identification of asystematic ageing management process which is an adaptation of Deming’s Plan-Do-Check-Actcycle to ageing management.

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Equipment Qualification (EQ) programmes provide an example of a potentially effective strategyfor managing of NPP components important to safety covered by these programmes (see ref. 11,IAEA list).

The objective of this report is to make available information on : EQ concepts and process, effectiveand practical methods and practices relating to upgrading and preserving EQ, and reviewing theeffectiveness of EQ programmes in operational NPPs. References are provided which direct readersto more detailed literature (e.g; : regulatory documents, standards, technical and programmaticguidance).

Ageing is particularly addressed in the following paragraphs :

2.4. Ageing and Qualified life (p 6)

3.3.2. Assessing ageing effects (p 29)

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The objective of this TECDOC is to document the current practices for the assessment andmanagement of the ageing of nuclear power plants steam generators (SG). It emphasizes safetyaspects and also provides information on current inspection, monitoring and maintenance practicesfor managing ageing of steam generators of the CANDU, PWR and WWER nuclear power plants.

Twelve countries have contributed to this document : USA, Belgium, Canada, Czech Republic,France, Germany, Japan, Russia, Slovenia, Spain, Sweden, Switzerland.

The document is structured in 8 technical sections :

Section2 : SG descriptions3 : SG design basis, fabrication and materials4 : SG degradation mechanisms5 : SG ageing management : operational guidelines6 : SG inspection and monitoring requirements and technologies7 : SG assessment methods and fitness for service guidelines8 : SG maintenance : mitigation, repair and replacement9 : SG ageing management programme

Recirculating steam generator tube (used in PWR) degradation mechanisms, discussed in Section 4,include (see fig.) : Primary Water Stress Corrosion Cracking (PWSCC), Outside Diameter StressCorrosion Cracking (ODSCC) (which includes both Intergranular Stress Corrosion Cracking(IGSCC) and Intergranular Attack (IGA), Fretting, Wear and Thinning, Pitting, Denting, High-Cycle Fatigue ; degradation process for SG shell, feedwater nozzle and tubesheet are : Corrosion-Fatigue, Transgranular Stress corrosion cracking, High-Cycle Fatigue, Erosion-Corrosion.

Operational Guidelines for SG ageing management described in Section 5 are grouped into thefollowing topic areas :

• Primary coolant system water chemistry control parameters,

• Secondary coolant system water chemistry control parameters,

• Measures to control secondary side impurity incursions

• Measures to remove secondary side impurities

• Measures to control SG deposits.

A very important aspect of SG ageing management is the use of a comprehensive inspection andmonitoring programme and appropriate fitness-for-service guidelines to assess the current andfuture safety of these components. These topics are discussed in Section 6 et 7.

Section 8 discusses mitigation and repair techniques for degradation mechanisms :

• in tubes :

184

− primary side : to mitigate PWSCC

- Rotopeening and Shot Peening

- Stress Relieving

- Reducing the Hot-leg Side Temperature

− secondary side : to mitigate IGSCC and IGA

- Control of the chemistry

- Reducing the Hot-leg Temperature

− tube repair : Plugging, Sleeving, Nickel Plating

− vibration control

• in feedwater nozzles and piping : modifications to minimize thermal fatigue.

The variety of maintenance actions available to manage ageing effects seems sufficient and thedocument does not suggest further development or research.

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Concrete Containment Buildings (CCB) need for systematic ageing management not only for thedesign life (30-40 years) or for the extended plant service life (60 years total being a quoted target)but also for the life after decommissioning if the containment is used as a "safestore" (up to 100years). Moreover considerations developed for CCB are of interest for other safety-related concretestructure.

The reports details the results of the Coordinated Research Programme (CRP) relating to the pilotstudy on concrete containment building (see paragraph 2.3) - Participating countries were :Belgium, Canada, India, Switzerland, UK, USA.

It addresses potential ageing mechanisms, age-related degradation, and ageing management (i.e.inspection, monitoring, assessment and remedial measures) for the following materials andcomponents for concrete containment buildings :

• concrete,

• reinforcing steel,

• prestressing systems,

• penetrations,

• liner systems,

• waterstops, seals and gaskets,

• protective coatings.

Structural steel piles and anchorages are also addressed to a limited extent.

The TECDOC does not address life or life-cycle management of CCBs because it is written fromthe safety perspective and life management includes economic planning.

The conclusions made are listed below.

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• The performance of the reinforced concrete structures in nuclear power plants has been good,with the majority of the identified problems initiating during construction and being corrected atthat time. However, as these structures age, incidences of degradation due to environmentalstressor effects are likely to increase the potential threat to their functionality and durability. Themost commonly observed form of degradation has been concrete cracking. Degradation factorsof primary concern would be corrosion of steel reinforcement due to carbonization of theconcrete or presence of chloride ions, excessive loss of prestressing force, excessivecontainment leakage due to failure of the metallic of nonmetallic boundary (lined) or excessiveconcrete cracking (unlined), and leaching of concrete.

• Techniques for detecting the effects of concrete ageing (i.e. inspection and performancemonitoring) are sufficiently developed to provide vital input for evaluating the structuralcondition of concrete containment buildings. Periodic application of these techniques providedata that can be used to trend performance and form the basis of an ageing managementprogramme. Areas of concern where these techniques require additional development include

186

massive members that contain large quantities of steel reinforcement, such as the basemat, andmembers that are inaccessible, such as portions of the steel pressure boundary that areembedded in concrete and foundations.

• Methods for conducting condition assessments or reinforced concrete structures are fairly wellestablished and generally start with a visual examination of the structure’s surfaces. Despite theusefulness of performing conditions assessments in maintaining the containment building’sfitness-for-service, with a few notable exceptions, the majority of the responses to the surveyquestionnaires sent to plant owners ant operators indicated that they do only limited evaluationsbeyond those mandated by codes and standards (e.g. visual examinations, leakage-rate testing,and prestressing tendon assessments).

Application of supplemental examinations and testing have primarily been associated withassessments of degradation occurrence or suspected occurrence.

• Maintenance and repair techniques for concrete structures are well established and whenproperly selected and applied are effective. At present no codes or standards are available forrepair of reinforced concrete structures, although some are being developed. Criteria that maybe used to determine when a repair action should be implemented are limited (e.g. parametersthat relate damage state such as crack with to environmental exposure). Data on the long-termeffectiveness or durability of remedial measures are required. Effective implementation of arepair strategy requires knowledge of the degradation mechanisms, the environment of thestructure at the macro and micro level, proper preconditioning of the structure to be repaired,correct choice of repair technique and material, and quality workmanship.

• Review of international practice has shown that many utilities worldwide have alreadyresponded to the potential for age-related degradation of CCBs and have implemented ageingmanagement programmes. These programmes generally adopt an approach in which any effectsof ageing are managed (as opposed to modifying operational environments to control theonset/rate of degradation).

• A characteristic of the most effective Ageing Management Programmes (AMP) was the cleardefinition and documentation of a systematic programme of activities aimed at understanding,effectively monitoring, and mitigating ageing effects. A particular feature was the routinetrending of surveillance and test data to estimate future performance of the CCB. This has valuein ensuring continued CCB’s fitness-for-service and hence plant availability.

• Drawing on international experience, a framework for ageing management of CCBs has beendefined (see figure). The proposed approach consist with existing IAEA guidelines. Anunderstanding of the issues involved (based on both plant specific knowledge and externalexperience on concrete behavior) is the basis for an effective AMP. The AMP consists of thefollowing key elements : (1) Definitions of the AMP to co-ordinate and integrate ageingmanagement activities and to identify the inspection and monitoring requirements andacceptance criteria ; (2) Operation of plant within design limits to minimize age-relateddegradation, in particular that which is error-induced ; (3) Inspection, Monitoring and ConditionAssessment to characterize significant component degradation before fitness-for-purpose iscompromised ; and (4) Maintenance to correct any unacceptable degradation (i.e. to manageageing effects). Technical guidance is provided for each of these tasks, together whit suggestedindicators for measuring the overall effectiveness of an AMP for CCBs.

187

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This report provides the technical basis for managing the ageing of the PWR and pressurized heavywater RPVs to assure that the required safety and operational margins are maintained throughoutthe plant service life. The scope of the report includes the following RPV components : vessel shelland flanges, structural weldments, closure studs, nozzles, penetrations and top and bottom closureheads. The scope of this report does not treat RPV internals, the control rod drive mechanisms(CRDMs), or the primary boundary piping used in PWRs. All the various size and types of PWRpressure vessels are covered by this report including the WWER plants builts in Rusia andelsewhere.

The technical information is mainly originated from US, Germany, Russia and France.

The structure of the document is as follows :

• Description of RPV

• Design basis : codes, regulations and guides

• Ageing mechanisms :

− radiation embrittlement

− thermal ageing

− fatigue

− corrosion

• Inspection and monitoring requirements and technologies

• Ageing assessment methods

• Ageing mitigation methods

• RPV ageing management programme.

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This report presents and discusses the requirements and methodologies utilized for the assessmentand management of ageing of PWR RVI (Reactor Vessel Internals).

This report provides the technical basis for managing the ageing of th PWR reactor vessel internalsto assure that the required safety and operational margins are maintained throughout the plantservice life. The focus of the report is on RVI components important to safety, however, forcompletness, RVI components not important to safety are also addressed in the report.

The structure is as follows : section 2 describes the RVI (Western and WWER types), including anoverall characterization of the design, importance to safety, materials and physical features of theRVI. In Section 3 , the applicable design basis, codes, standards and regulations are addressed.Section 4 deals with operating conditions, Section 5 identifies dominant degradation mechanisms(embrittlement ; fatigue ; corrosion ; radiation induced creep ; relaxation and swelling ; mechanicalwear), sites, consequences, and significance of degradation mechanisms. Section 6 addresses theapplication of inspection technology to assess the condition of the RVI. Section 7 summarizes thecurrent knowledge on service experience and related maintenance. Section 8 describes an ageingmanagement programme for PWR RVI utilizing a systematic ageing management process andoutlines relevant national and international ageing research.

The technical information is mainly originated from USA, Germany, Japan, Russia and France.

189

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The CRP on management of ageing in-containment I&C cables was initiated at the first ResearchCo-ordination Meeting (RCM) held in Vienna in December 1993. The general objective of the CRPwas to identify the dominant ageing mechanisms and to develop an effective strategy for managingageing effects caused by these mechanisms. The specific objectives were :

(a) to validate prédictive cable ageing models accounting for synergistic effects that take placewhen radiation and thermal ageing occur over the long time period associated with real plantenvironments, and

(b) to provide practical guidelines and procedures for assessing and managing the ageing ofI&C cables in real plant environments.

The scope of the CRP was limited to those materials and cables types which were considered to beof widest interest. The programme was therefore limited to low voltage (< 1 kV) I&C cables basedon crosslinked polyethylene (XLPE), ethylene propylene based materials (EPR/EPDM) aandethylene vinyl acetate (EVA). Because of their similarity in materials and construction, low voltagepower cables were also included in the programme.

The CRP was implemented in two phases. Result of Phase I CRP (1993-1995) are presented inIAEA-TECDOC-932 "Final report : Pilot Studies on Management of Ageing of Instrumentation andControl Cables". They include a summary of the relevant ageing mechanisms ; operating experiencefor a range of NPP types ; an overview of ageing management methods which are currently in use ;description of cable sampling and laboratory ageing methods and of monitoring and test methods ;the capabilities and the limitations of the various ageing management methods.

The objective of Phase II CRP was to resolve the uncertainties in the relationship between cablecondition monitoring techniques and DBE survivability and improve existing initial qualificationprocedures, and thus to provide a technical basis for the assessment and management of ageing in-containment I&C cables based on the concepts developed in Phase I CRP. Most of the CRP effortwas aimed at : the identification of cables of concern in order to focus limited ageing managementresources on a manageable sub-set of the total cable inventory in plant ; developing a 'tool box' ofpractical condition monitoring (CM) methods trough round-robin tests (to identify the most suitableCM methods and their limitations for different cable materials and applications, and to develop testprocedures for these methods) ; and the correlation of CM methods with the survivability of cablesin a DBE.

Since PVC cables are used in many of the existing NPPs, they have been included in the scope ofPhase II.

A final report "Assessment and management of ageing of in-containment I&C cables : Technicalbasis" is planned for mid 2000.

Countries participating in the CRP are : Canada, Czech Republic, France, Germany, India, Japan,Romania, Russia, Sweden, Switzerland, UK, USA.

190

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This Safety Report complements the Safety Guide No 50-SG-02 "Periodic Safety Review ofOperational Nuclear Power Plants - It provides practical information for the Safety Assessment andjudgment process, and a methodology for ranking safety issues identified by the review.

In addition this report provides assistance on the prioritization of correctives measures and theirimplementation so as to approach an acceptable level of safety.

It follows the recommendations of the INSAG-8 documents ("A Common Basis for Judging theSafety of Nuclear Power Plants Built to Earlier Standards" - IAEA, 1995).

Several factors have to be considered in the scope of a complete safety review and the managementof ageing is one of them.

191

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This publication establishes the requirements that must be met to ensure the safe operation of land-based stationary thermal neutron power plants and also includes this commissioning and subsequentdecommissioning.

It does not contain specific requirements relating to ageing.

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This Safety Guide provides guidance on the conduct of PSRs for an operational nuclear powerplant. The Guide is directed at both owners/operators and regulators.

Ageing aspects are dealt with in paragraphs 416 to 419 : Equipment Qualification and mainly inparagraphs 420 to 423 : Management of Ageing. It gives only general statements on this aspect andrefers to specialized documents issued by IAEA ("Methodology for the Management of Ageing ofNPP Components Important to Safety" Technical Report Series No 338 and "Data Collection andRecord Keeping for the management of nuclear power plant ageing" Safety Series No 50-P-3).

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The purpose of this Safety Guide is to provide guidance on the conduct of PSRs for an operationalnuclear power plant. It is directed at both owners/operators and regulators.

PSRs are undertaken about every ten years and during this period significant changes may occur,among which are the consequences of ageing, and must be taken in account in order to obtain anoverall view of actual plant safety.

The management of ageing is dealt with in the paragraphs 420 through 423 and in the appendix onElements of Review.

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This report represents a review of published information related to three cost categories that are partof cost-benefit analysis : costs upgrades necessary for continued operation of a nuclear unit, costs oflifetime extension measures, and costs of decommissioning. While each of these categories issubjected to detailed specialised cost studies, the report views the costs globally, mainly as input forsubsequent overall economic analysis. Consistently with this approach, the report also discusses theapplicability of the collected costs for decision making.

Some emphasis is placed on the so-called "Soviet-designed" reactor models (WWER 440 and 1000; RBMK 1000 and 1500).

The cost data presented are useful as information suitable for general understanding of themagnitude of the costs but, due to the large variability of the costs (by country, by model, byvintage ...), they cannot be used for a complete analysis of the economics of continued operation ofolder NPPs.

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The pilot studies, initiated in 1989, are being implemented in two phases using the methodology forageing management studies in TRS N° 338. This TECDOC gives the result of Phase I studies(interim ageing studies) ; the next step (Phase II) deals with comprehensive ageing studies,implemented through IAEA co-ordinated research programmes (CRPs).

The work performed in Phase I consisted of a review of current understanding of ageing andmethods for monitoring and mitigation of this ageing for the four selected NPP components :

• primary nozzle of a reactor pressure vessel,

• Motor Operated isolating Valve (MOV),

• concrete containment building,

• instrumentation and control cables within the containment.

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Two types are considered :

• the RPV inlet and outlet nozzle region of a PWR,

• the RPV feedwater nozzle region of a BWR

for which are described the specifications, standards and regulations, the design and fabrication ; thepotential degradation mechanisms and operating experience ; and monitoring techniques.

The knowledge and technology gaps identified are mainly :

• understanding of environmental effects (oxygen contents and flow rates) on corrosion fatigue,

• measurement of the fatigue damage (no fully validated method),

• effectiveness and reliability of inspection techniques for crack detection,

• verification of assessment methods

− conservatism of fatigue design curves based on small scale specimens,

− effect of stress concentration,

− distribution of residual stresses,

− demonstration of the conservatism of simplified approaches and models ,

− development of probabilistic failure analysis.

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A rather qualitative overview is presented on the general background with the same contents as thepilot study on RPV primary nozzle.

The knowledge and technological gaps identified are as follows :

• understanding of MOV ageing : development of a database on the failure and malfunction inplant operation,

• monitoring of MOV ageing : development of system of device,

• risk and reliability assessment : taking into account of the degraded MOV conditions andfailure,

• improvement of qualification methods,

• improvement of maintenance procedures.

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After a general presentation the current understanding of ageing of concrete containment building isdetailed (concrete / reinforced steel / prestressing system / liners and penetration / waterstop).

The knowledge and technology gaps pointed out are :

• current experience and ageing management practices,

• state-of-the-art repair techniques,

• crack mapping and depth measurements,

• conditions indicators.

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Knowledge and technology gaps relating to significant service conditions that can influence cableageing are :

• effects of temperature : well understood but practical in-situ methods for thermal degradationnot currently available,

• lifetime assessment necessary for cables that face total dose more than 30 Kgy,

• quantification of the effect of oxygen content in a radiation environment.

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This report is focused on the proper management of the operating lifetimes of nuclear power plantsin order to help to minimise or eliminate avoidable or premature plant shutdowns. It complementsthe INSAG-8 report “A common basis for Judging the Safety of Nuclear Power Plants Built toEarlier Standards”. It covers mainly the treatment of physical ageing of structures and componentsin order to maintain operating plants consistent with their initial design basis but it also considersthe infra-structural issues that influence the capability to provide adequate means for safemanagement of operating lifetimes of nuclear power plants (maintaining adequate competence,handling major organisational changes).

It promotes safety reviews to provide an overall view of the actual safety status of a plant(effectiveness of the ageing management ; discussion on the possible evolution of the referencesafety level).

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The various parameters considered for the prioritisation of components, taking into account safetyand economic aspects appear to be similar in the three countries. The main difference betweenpractices is on the respective weights placed on these various parameters, depending on utilitiesconditions and strategies.

The classifications of systems, structures and components evaluated for the three countries aregiven in chapters 2, 3 and 4 for Belgium, France and Spain respectively.

In order to illustrate more precisely how the various aspects may be considered in order to get aglobal classification of SSCs, the screening methodology developed by UNESA is given below.

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Plant structures, equipment, hardware, and components are compared to a common set of weightedevaluation criteria to identify those most important to the lifetime of the plant. These criteria andweights have been sensitised to reflect technical and non-technical issues affecting nuclear powerplants. For example, technology for establishing the effectiveness of plant programs to manageageing mechanisms has been applied.

Six criteria categories are used: General Criteria, Service Conditions, Regulatory Factors, ServiceHistory, Plant Programs Effectiveness, and Reliability Considerations. Weighting factors are usedto properly characterise the relative importance of criteria categories, and individual criteria, withrespect to their threat or benefit to the lifetime of the plant. The category entitled “plant programseffectiveness” was added to previous screening criteria to factor in the importance of maintenanceand ageing management programs to achieving lifetime related goals.

Plant structures and hardware components are selected using the pre-screening process discussed inSub-section 1.2.1. The pre-selected components are then organised into one of 13 groupings aslisted in Table 1. Each of these groupings contains structural or hardware items, which may have animpact on plant life extension.

1.2.1. Candidate Selection Process

A complete criteria screening or evaluation for all of the structural and hardware items or types in aLWR plant would be a very major undertaking. A pre-selection process is therefore applied togenerate a manageable listing of items to be included in the important components evaluation. A listof components important to Lifetime Cycle Management is obtained in two phases:

• Selection of systems

• Selection of components

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1 Reactor Vessel Components2 Primary Containment Components3 Primary Concrete Structures4 Primary Steel Structures5 Secondary and Outbuilding Structures6 Pumps and Turbines7 Tanks and Heat Exchangers8 Piping Systems9 Valves10 Other Major Mechanical Equipment and Components11 Primary Electrical Equipment and Components12 Major Electrical Motors13 Primary I&C Equipment and Components

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A system or structure will be considered important to Lifetime Management whenever it falls underone of the following categories:

• Nuclear Systems: Systems that monitor or control core radioactivity, remove heat from the coreor are directly involved in the safe operation of the reactor

• Safeguard Systems: Systems (excluding containment systems) that are used to mitigate reactoraccident.

• Containment Systems: The containment and systems that prevent its overpressure or excessleakage through the containment to the environment

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A system or structure will be considered important to Lifetime Management is it is essential to theproduction of electricity and does not have redundancies with respect to its basis functions andtherefore, its failure would have a significant impact on power plant availability.

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Systems important to Lifetime Management will be considered those systems with at least oneelement of high unit cost or difficult to replace (due to long shutdown period or procurementdifficulties) and those structures whose reconstruction is not foreseen.

Whenever a system or structure meets any of the above criteria, it will be considered potentiallyimportant to Lifetime Management.

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Once the systems important to Lifetime Management have been defined, their components will beevaluated using a group of selection criteria to determine the components that are important toLifetime Management. The components on the resulting final list will be ranked by their importancefor Lifetime Management.

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Listing system components according to their individual identifiers the plant would be lead to alaborious selection process and an excessively long and repetitive list inappropriate for setting

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priorities for study, evaluation and effort in the plant LCM, which is the final objective of listingImportant Components. To avoid this, before selecting LCM-important components, the selectionand ranking processes must be optimised by defining the components of each system according todiscretization or scope criteria. For example components of mechanical system will be listed as:

− Main equipment. Components that perform fundamental system functions (pumps,exchangers, large filters, tanks, evaporators, etc)

These will be listed according to their function. Redundant equipment will be named inplural without specification of the trains to which it belongs (ie, RHR Pumps includes theRHR pump-train A, RHR pump-train B and RHR Pump-train C)

− Piping: Listed as a generic component

− Valves and Dumpers: Listed according to the following types: motorised, manual, check,solenoid and Safety and Special

− Instrumentation: Listed as a generic component

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Once the Components Lists of LCM-Important Systems have been prepared, these will be selectedaccording to the following criteria:

1. The component, if it were to fail, would have a significant impact on plant safety andavailability

2. The component is of special and specific relevance to the plant licensing process

3. The component is safety-related or required for the cold shutdown of the plant and itsmaintenance

4. The operation or ambient conditions of the component are more aggressive than thoseconsidered in design

5. The component has required a significant maintenance effort during its lifetime

6. Maintenance of the component is not considered effective for the control and/or mitigation ofageing

7. Replacement of the component implies a large expense, associated with the cost of thecomponent itself or the extended period of plant shutdown

Fulfilment of any of the above criteria means that the component is important to LifetimeManagement.

1.2.2. Screening Methodology

The screening methodology uses a systematic scoring process to establish a "total criteria score" foreach of the candidate structural and hardware components. The screening process provides amethod of quantifying an otherwise subjective selection process. Comparison of the total scores willlead to the determination of the important structures, equipment, hardware, and components. Thedominant factors that influence the selected important components can be identified by reviewingthe criteria scoring.

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General Criteria Service Conditions Regulatory Factors Service History Plant Program Effectiveness Reliability Considerations

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This section provides definition for each criterion to be listed in the component matrices.

1.3.1. General Criteria

Five criteria are grouped in this general criteria category. These items deal with the discrete issuesof replacement feasibility, plant outage duration, replacement cost, impact on adjacent plantstructures and hardware, and case of disposal and transportation. As a group, these criteria will havea significant impact on the feasibility of achieving lifetime goals.

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Replacement means total removal of a structure or hardware item from its current location. Theitem would then be moved to a new location and restored to service, replaced by a new item of thesame design, or replaced by a new item with a new design. The feasibility of replacing a structure orhardware item is established by considering the availability of a qualified replacement component,accessibility limitations, construction, radiation exposure levels, etc. Outage length, costs, disposalof contaminated wastes, and the effect on adjacent components are addressed in other criteria.

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Outages are defined as the scheduled or planned shutdowns for refuelling, or for the otherreplacement, modification or inspection/surveillance reasons. It is fully expected that plannedoutages will be scheduled at refuelling periods. As such, the optimum outage length would be anormal refuelling outage duration. LCM program activities, which do not affect the normalrefuelling outage critical path, would be acceptable. If the activity is expected to increase the outageduration, then an assessment of whether the outage length is tolerable is required. The aspectsaffecting this assessment would be the system reserve generating capacity, the availability factors ofthe other plants in the system, and the purchased power capacity/availability. In addition, theongoing fixed plant costs (salaries, property taxes, etc.) and any lost revenue would be considered inthe outage length assessment. The direct costs associated with the LCM activity and thereplacement power costs are considered in the Subsection 1.3.1.3 criterion.

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The term “replacement” is defined in Subsection 1.3.1.1. The cost magnitude of a structural orhardware replacement is to be weighed against the current and future cost to replace the entire plant.Structural or hardware replacement total costs in the range of 40% to 50% of the entire plantreplacement cost could subjectively be considered a prudent upper limit when evaluating thiscriterion. All replacement costs other than outage costs are covered by this criterion. Outage cost isa consideration of Item 1.3.1.2, "Outage Duration".

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Adjacent structures and components can be generally categorised as direct attachments (that is,service water, air or electrical connections, and supporting structures, components or foundations)which must be removed or disconnected to perform the desired work. The work activities wouldinclude structural or hardware replacements, modifications, or enhanced maintenance andinspections. Structures, equipment or components, which must be removed to establishaccessibility, material staging areas, and hardware removal and installation paths, would also beconsidered in this category. Other influencing factors would include lay-up, decontamination, andshielding requirements for adjacent structures and components.

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Replacements, modifications, decontamination, or destructive testing activities may lead to thegeneration of radioactive or hazardous wastes which pose special disposal and transportationproblems. When considerable volumes or high level waste materials are involved, special transportequipment and onsite or offsite storage structures may need to be built. In the years to come, theability to fully off-load the reactor core may be limited by the accumulation of spent fuel in thestorage pools. Provision for long-term dry storage or temporary storage provisions may have to bedesigned and licensed to store activated materials or spent fuel.

1.3.2. Service Conditions

The six criteria described in this subsection acknowledge the impact on the potential remanent lifeof a component due to service conditions. A candidate structure or hardware item which mustfunction under known adverse conditions is more likely to limit the lifetime than a comparablestructure or hardware item operating under less severe conditions. The criteria are presented interms of possible failure modes associated with the service conditions. The weighting factorsassigned to these criteria reflect the uncertainty and severity of the possible failure.

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Failure in a non-ductile manner generally occurs without warning, and is usually catastrophic.Protection against brittle fracture is an important aspect in the design, construction, and operation ofmany components. However, conditions that increase the likelihood of brittle fracture includesignificant loading near the nil ductility temperature (NDT). Other conditions that tend to increasethe NDT include exposure to significant neutron flux levels (Section 1.3.2.4) or strain hardening.Thermal embrittlement of cast stainless steel components and other materials is also included in thiscategory.

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Components that are subjected to corrosive environment will have obvious lifetime concerns. Theincreasing observation of microbiological corrosion in nuclear power plant water bearing systemsand in other components suggests that it is an important factor to consider for long-termperformance.

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Components, which are subjected to significantly changing loads over time generally, will have ashorter life than those subjected to only static loads. These conditions include suddenly appliedloads such as those due to fast valve closure or opening, water hammer, or impact. Fatigue fromthermal cycles or transients may also occur if not considered in design (or if design provisions havebeen exceeded).

Components operating under cyclic or fluctuating loading conditions may be susceptible to variousfatigue mechanisms. A distinction can be made between components subject to low cycle fatigue

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(approximately 10,000 cycles or less) and those subject to high cycle fatigue (greater than 10,000cycles). Generally, low cycle fatigue conditions are characterised by high amplitude load (stress,strain or strain rate) and shorter lifetimes.

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Materials that are susceptible to radiation damage and located in a radiation field will have obviousservice life concerns. Exposure to nuclear radiation may change the properties of certain componentmaterials, such that the item cannot perform its design function. Failure modes associated withradiation damage generally include a loss of ductility and degradation of electrical or electronicperformance. Loss of ductility sometimes enhances the strength properties of a material, but maylead to brittle failure as described in Subsection 1.3.2.1. For some materials, such as concrete andelectrical cable, the effects of radiation on service life are not completely known. Irradiation effectsare especially important to PWR reactor vessels.

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Certain materials, such as sensitised stainless steel, in various environments and stress fields, aresusceptible to stress corrosion cracking (SCC). Of particular concern are systems and componentsexposed to reactor coolant water or fluids containing resins or chlorides. The potential for SCCincreases for components of welded fabrication, exposed to temperatures exceeding 100ºC, andunder a continuous tensile stress field. SCC is particularly important, because it generally leads to afast fracture risk in a normally ductile material.

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Environmental exposure to the flow of fluids containing solids or vapour bubbles, other movingobjects, or aggressive chemicals may lead to a loss of component material or cracking. Often times,these mechanisms occur over a fairly long period of time and display visible signs of material lossbefore reaching critical stages. Other times, such as during a chemical spill, flood or intrusion ofsolids / particles, the rate of attack may be enhanced. Within composites such as thermoplastics andconcrete, these events and others also lead to the formation of cracks and loss of function.

1.3.3. Regulatory Factors

The three criteria described in these subsections acknowledge the potential impact that regulatoryconsiderations or changes will have in lifetime evaluations and planning.

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Lifetime program evaluations will need to emphasise the systems, structures, and componentswhich were key considerations during original plant licensing and subsequent updates. This wouldinclude balance of plant hardware whose failure could result in challenges to safety systems, orcontribute to processing of radioactive waste.

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The operability and/or functionality of components that are required to safely shut down the plantduring transient or accident conditions will be a key of lifetime evaluations and justifications. Thestructures, systems, and components involved are a subset of the items discussed in Subsection1.3.3.1. Other factors such as redundancy, separation, and single failure assumptions are alsoimportant considerations for safe shutdown hardware. Therefore, lifetime considerations will likelyaffect many of the safe shutdown hardware types.

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The performance of certain systems, structures, and components, or lack thereof, has significantimpact on the function of emergency core cooling systems (ECCS). Components whose failure

204

could result in loss of reactor coolant, reduced performance of an ECCS operating component (i.e.,RHR pump bearing cooling, etc.), or would initiate an ECCS are important to evaluation in LCMprograms.

1.3.4. Service History

The three criteria described in this subsection consider the factors, which characterise theperformance of the structure or hardware item to date, and its potential remnant life.

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Hardware items that have historically displayed a low availability and a high maintenancefrequency will also have questionable life expectancies. The maintenance frequency will mostprobably be indicative of a severe environ-ment, severe cyclic duty, or an unavoidable materialapplication problem. Should it be established that the hardware item is required for safe shutdownand is a important element in the LCM program, extensive evaluations and testing may be necessaryto develop a remedy and to support the lifetime goal.

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The in-service inspection programs include visual, surface, and volumetric examinations of pressureboundary materials. Other plant surveillance and inspections programs examine the condition ofelectrical, instrumentation, steel structures, foundations, etc. Defects, flaws or deviations identifiedduring these examinations are subjected to evaluation criteria. The flaw or deviation may requirecorrective action including repair, or may be dispositioned as acceptable. Those flaws or deviations,which are found acceptable, are periodically monitored to ensure that the situation is not becomingworse. The known existence of flaws or deviations in a important structural or hardware item willundoubtedly lead to additional efforts to support lifetime goals. If a repair is not desirable or costeffective, then it will be necessary to rely upon analysis or experimental techniques to establishadequate safety margins.

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Previous regulatory, design or performance factors may have already led to a full or partialreplacement of a structure or hardware item. If this is the case, it may be that the replacement has alife expectancy that exceeds the current service life goals for the plant. This will be a benefit to theLCM program. It is also possible that the replacement has a life expectancy the same as the originalcomponent. This will be of some benefit in the LCM program because of the experience gainedfrom replacement activities (component improvements, etc.).

1.3.5. Plant Programs Effectiveness

The four criteria described in this subsection address the availability of plant programs and theireffectiveness in monitoring, maintaining, and preserving the function of important plantcomponents.

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The presence, availability, and usage of formalised maintenance, inspection, and surveillanceprocedures have a definite and positive impact on achieving remanent life goals. Plants, which haveformalised procedures containing measures to mitigate ageing effects at early stages via preventiveaction will enjoy increased operating efficiency and reduced outage time. The importance of thiscriterion to long-term operation is noted as a weight of seven (7) has been assigned.

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Certain plant procedures and programs contain elements or steps specifically to ensure that thesubject system, structure, or component performs suitably in the future. Entitled

205

“preventive/protective measure”, these activities are often elements of a formalised program(Section 1.3.5.1) and supplement normal or corrective maintenance.

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An additional element of prudent preventive maintenance and surveillance procedures is acceptancecriteria upon which future performance is tied to. During the maintenance/surveillance activity,certain performance parameters (quantitative) are obtained and compared to a specified criteriarange or value. If the criterion is exceeded, provisions to take corrective action are employed. Useof this type of activity results in increased effectiveness of maintenance practices and reducedcomponent down time.

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Preservation of key operating and maintenance history documents for systems, structures, andcomponents has been found to be very important to technical justification of remaining life.Formalization of this activity should be a part of plant surveillance, maintenance, and otherprograms. The records should be retained in a method that is non-aggressive each medium type,protected from a natural hazard(s), and allows easy retrieval and usage. Similarly, it is importantthat the, records retained have the proper technical data needed for lifetime assessment.

1.3.6. Reliability Considerations

The five criteria described in this subsection address methods that are indicators or influencecomponent reliability.

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Systems, structures, and components, which have a demonstrated track record of suitableperformance, reliability, and availability, provide greater assurance that long-term plant operation isfeasible. Conversely, components that are commonly out of service, or require constant attention orhave suspect performance, have a reduced potential for long-term operation. This criterion isespecially important to components having an "active" function. A credible service life of acomponent is very important to justifying long-term function.

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Systems, structures, and components whose malfunction may cause a forced outage of the plant areconsidered important to life cycle management. This criterion refers to safety-related SSC and anyBOP components such as primary structures with important functional requirements.

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The performance and function of certain plant systems, structures, and components (SSC) has beenpreserved by the addition of back-up or support systems and procedures. Existence of supplementalsystems and procedures to improve performance or mitigate ageing, are attractive measures from aservice life perspective. Redundancy to achieve safety requirements should not be considered in thiscontext.

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In several areas of manufacturing, fabrication or support services, the number of qualified suppliershas been declining. The term "qualified" applies to the quality assurance, environmentalqualification, code stamping, etc., requirements for structural and hardware items supplied to thenuclear industry. As the number of plants under construction continues to decline, some vendorshave not maintained the required certifications, and have discontinued their product lines. As such,it may be difficult to replace or repair some structures or hardware items.

206

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In addition to the cost for replacing certain components (Section 3.1.3), there are reliabilityconcerns for finding and obtaining replacement parts to keep components operational. As thenuclear industry increases in age without new nuclear generation capacity added obsolescence couldcause significant difficulty in locating replacement parts. Already, components such as the BWRcontrol rod drive hydraulic control units require significant maintenance, and spare parts aredifficult to obtain. Without new plant orders, the timing to obtain such parts may introduceconsiderable economic penalties.

1.3.7. Other Considerations

Space has been provided on the matrix form to include other criteria should it be necessary duringthe evaluation process. Such additional criteria should be discrete. That is, they should be evaluatedto determine whether or not already established criteria (Sections 3.1 through 3.6) can be applied. Ifthis option is selected, the criterion and the basis for scoring shall be described and included in theevaluation records. Appropriate weighting factors will also have to be established and entered onthe table. Care shall be exercised not to duplicate already established criteria.

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The screening procedure described in Section 1.3.2, combined with the established criteria ofSections 1.3.1 to 1.3.7, provides a suitable method for determining the relative sensitivity ofsystems, structures, and components important to ageing and the lifetime. These upgraded screeningcriteria also address the importance of maintenance and may be applied to BWR or PWR plants.

Following use of this criteria document, a listing of important plant components is obtained forfurther technical review and given in Section 4 of this Appendix. Future tasks within the Life CycleManagement Program will define the requirements for these reviews, the technical issues andageing concerns, and preparation of "licensing basis” documentation. The following Lists are theresults of UNESA methodology application in two pilot plants (PWR and BWR). Other applicationwill provide different results.

207

�� %(/*,80�$3352$&+The table below gives the result of the general ranking tentative of existing Equipment AgeingSummaries.

6\VWHPV��6WUXFWXUHV��DQG�&RPSRQHQWV 3UREOHP�FDWHJRU\5HDFWRU�SUHVVXUH�YHVVHO Core belt region Irradiation embrittlement&RQFUHWH�VWUXFWXUHV Seismical category 1 Leak tightness6WHDP�JHQHUDWRU Corrosion, wear, vibration and fatigue5HDFWRU�FRROLQJ�SXPS Thermal barrier Thermal fatigue5HDFWRU�SUHVVXUH�YHVVHO Cover head penetrations Primary water stress corrosion cracking3LSLQJ Stress and fatigue due to unexpected

transients3ULPDU\�FRQWDLQPHQW Pre-stressed cables Loss of prestressing3UHVVXULVHU Surge line nozzle Cracking (Fatigue, Primary Water SCC)9HVVHO�LQWHUQDOV Wear3UHVVXUH�YHVVHO Bottom mounted instru-

mentation penetrationsPrimary water stress corrosion cracking

3LSLQJ Flow accelerated corrosion,(�HOHFWULFD��DQG�,&HTXLSPHQW

1 E qualification programs andpreventive maintenance

(OHFWULFDO�FDEOHV Cable ageing5HDFWRU�SUHVVXUH�YHVVHO Canopy seal welds Stress corrosion cracking3LSLQJ Micro biological corrosions9HVVHO�LQWHUQDOV Baffle bolts Irradiation assisted stress corrosion

cracking&RQWURO�URGV Flow induced fretting wear$60(�6HFWLRQ�,,,��'LY�� $OO�Class 1 components Stress and fatigue9HVVHO�LQWHUQDOV Guide tube split pins Stress corrosion cracking of Inconel 7502WKHU�VWUXFWXUHV Degradations due to concrete

carbonation3ULPDU\�SLSLQJ Cast stainless steel elbows Thermal ageing5HDFWRU�FRROLQJ�SXPS Cast stainless steel casing Thermal ageingPiping Stress and fatigue due to regulatory

reconciliation3UHVVXULVHU Heating rods Primary water stress corrosion cracking

208

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The table below gives the general ranking of Equipment obtained according to Frenchmethodology.

0$5. /,67�2)�&20321(176�7$.(1�,172�$&&28171 Reactor Pressure Vessel2

Life Time ProjectContainment

3 High diameter primary piping4 Class 1, 2 and 3 piping5 Steam generators6 Pressurizer7 Reactor Internals8 Instrumentation and control, Converters9 Electric cables

10 Turbine11 Generators12 Primary coolant pump13 Control rod system14 Vessel supporting structure15 Anchor bolts16 Cooling towers17 Handling facilities18 In-core flux thimble19 Bimetallic connections20 Alloy 600 areas21 Charging pumps, SG feed pumps22 Nuclear heat exchangers23 Condenser24 Turbine heater & separators-reheaters25 Circulation pump speed reducer26 Feed pump27 Emergency diesels28 Transformers29 Civil works for nuclear island30

Current Maintenance andExceptional Maintenance

Valves

209

�� 63$1,6+�$3352$&+

The following lists are example results of applying the UNESA methodology to two pilot plants(PWR and BWR). Other applications would provide different results.

3:5

LISTING AND RELATIVE RANKING OF COMPONENTIMPORTANT FOR PLANT LIFE MANAGEMENT

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1 Reactor Pressure Vessel 5882 Steam Generators 5613 Reactor Building 5404 Reactor Vessel Pedestal 5345 Metal Primary Containment 5316 Auxiliary Building 5207 Reactor Coolant Pumps Foundation 5148 Pressurizer Pedestal 5129 Steam Generators Pedestal 51210 Essential Service Water Piping 49611 Primary Containment Mechanical Penetration Assemblies 46512 Pressurizer 46313 Intake Structure 46214 Primary System Equipment Supports 46215 Electrical Building 46116 Fuel Building 45517 Turbine Peetration Building 44918 Componet Cooling Building 44119 Diesels Building 43420 Reactor Coolant Piping 42821 Diesel Generators Fundations 42322 Cables inside Primary Containment 41523 Essential Service water Pumps Foundations 41424 Turbine Bulding 40925 Fuel Pool Liner 40526 RHR Pumps Foundations 40427 Charging Pumps Foundations 40028 Turbine Foundation 39929 Reactor Coolant Pumps 39830 Fuel Transfer Tube 39631 Essential Diesel Generator (Engine) 38332 Emergengy Diesel Generator (Engine) 38333 Polar Crane 37834 Cables outside Primary Containment 37835 Main Turbine 37536 Intake Electrical Building 37537 Contaiment Spray Pumps Foundations 37338 High Pressure Safety Injection Piping 37139 Snubbers 369

210

40 RHR Piping 36841 Primary Containment Personnel Airlock and Equipment

Hatches368

42 Component Cooling Heat Exchangers Foundations 36743 Primary Containment Electrical Penetration Assemblies 36444 Control Building to Auxiliary Building 36045 Reactor Pressure Vessel Internals 35846 Radwaste Tunnel 35747 Diesel Generators (Generator) 35648 Main Steam Isolation Valves 35049 Safety Inyection Piping 34850 Main Steam Piping 34751 Component Cooling Heat Exchangers 34652 Electrical Boxes Inside Contaiment 34653 Charging Pumps 34054 Main Steam and Feedwater Supports 34055 Diesel Generatos Electrical Cabinet 33756 RHR Pumps 33457 Containment Spray Piping 33458 Essential Service Water Pumps 33459 Feedwater Control Valves 33360 Main Condenser 33061 Bateries 32762 Auxilary Feedwater Piping 32563 1E Distribution Centers 32364 1E Motor Control Centers 32365 1E Load Centers 32366 Control Rod Driver 32267 Refuelling Water Tank 32168 Main Steam Safety and Relief Valves 32069 Electrical Boxes Outside Contaiment 31970 Safety Inyection Accumulators 31471 Neutron Flux Detectors 31472 24V Power Supplies 30673 Condensate Tank 30574 Component Cooling Pumps 30375 Low Voltage Motors 29976 DC Load Centers 29777 Boric Acid Tank 29678 Auxilary Feedwater Pumps 29679 Main Turbine 29580 Liquid Radwaste Tanks 29581 1E Instrument Load Centers 29382 1E Instrument Auxiliary Load Centers 29383 Primary Containment Isolation Valves 29184 1E Intrument Inverter Power Supply 28985 Main Generator 28986 Reactor Coolant Drain Heat Exchanger 28887 6,3 kV Motors 28888 Main Transformers 286

211

89 Radwaste Monitoring Tank 28390 Safety Inyection Accumulator Check Valves 28291 No 1E Motors Control Centers 28292 Main Turbine Piping 28293 Motor Operated Valves 28294 No-1E Distribution Centers 28295 No-1E Load Centers 28296 Auxiliary Feedwater Turbine 28097 Bateries Chargers 27898 Refuelling Handling Crane 27899 Auxiliary Feedwater Pump 277100 Radiation Monitors 275101 Volume and Chemical Control Piping 274102 Steam Extraction, Drain and Vent System Piping 273103 Boron Recycle System Tanks 271104 Refuelling Building Crane 270105 Essential Chillers 268106 Cathodic Protection 268107 Instrument Cabinets (electronic cards) 268108 Containment Low Capacity Filtering Unit 267109 Auxiliary Feedwater Tank 267110 Auxiliary Turbines Hydraulic Valves 265111 Auxiliary Transformers 263112 Reactor Coolant Drain Tank 263113 Volume Control Tank 262114 Equiment Drains Tank 261115 Circulating Water Piping 261116 Grouped Phase Buses 260117 Essential Chilled Water Pump 255118 Special and Instrument Cables 254119 Containment Sum Grids 253120 400kV, 220kV, and 110kV Swtichgears 251121 Circulating Water Pumps 250122 Control Rod Driver Motor-Generator Sets 250123 Hydrogen Recombiner 243124 Circulation Water Movile Grids 240125 Mix Bed Desmineralizer 239126 Cathionic Bed Desmineralizer 239127 Main Generator Swtichgear 238128 Electrical Protection 237129 EDG Gas-oil Storage Tank 236130 Containment Blowdown Low Capacity Unit 236131 Heaters Drain Tank 235132 Reheater and Moisture Separators 235133 Transmisor Elements 234134 Fuel Pool Desmineralizer 234135 Isolated Phase Buses 234136 Reactor Make-up Piping 232137 Acid Boric Pumps 231138 Regenerative Heat Exchanger 228

212

139 Refuelling Water Heat Exchanger 228140 Temperature Elements 228141 Pressurizer Relief Tank 226142 Excess Let Down Heat Exchangers 224143 Let Down Heat Exchangers 224144 Containment High Capacity Filtering Unit 223145 Containment Spray Pump 221146 Plant Computer 220147 Refuelling Water Storage System Piping 218148 Fuel Pool Heat Exchanger 217149 Steam Generator Blowdown Electromagnetic Filter 216150 Limit Swtiches 215151 Seal Water Heat Exchanger 214152 Reactor Make-up Storage Tank 214153 Steam Generator Blowdown Piping 213154 Containment Low Capacity Blowdown Ventilator 211155 Air Compressors 210156 High Pressure Feedwater Heaters 208157 Low Pressure Feedwater Heaters 208158 Fire Protection Water Tank 205159 Fire Protection Diesel 205160 Feedwater Pumps 205161 Containment Blowdown High Capacity Unit 203162 NaOH Pump 196163 Auxiliary Feedwater Hydraulic Valves 196164 Feedwater Turbines 195165 Solenoid Valves 194166 Essential Service Water Auxiliary Storage Tank 185167 Desmineralized Water Tank 185168 Feedwater Check Valves 183169 Containment High Capacity Blowdown Ventilator 183170 Seals Cleaning Water Storage Tank 177171 Air Compressed Post-Cooler 171172 Air Compressed Dryer 168173 Gas-oil Transfering Pumps 167174 Fuel Pool Cooling Pumps 167175 Main Turbine Control and Stop Valves 166176 Air Operated Valves 165177 Fire Protection Pump (Jockey) 150178 Fire Protection Pump ( Electrical) 148179 Fire Protection Pump (Diesel) 148

213

%:5

LISTING AND RELATIVE RANKING OF COMPONENTIMPORTANT FOR PLANT LIFE MANAGEMENT

5DQN &RPSRQHQW�6WUXFWXUH�&RPSRQHQW�*URXSLQJ�'HVFULSWLRQ

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1 Reactor Pressure Vessel 6372 Reactor Vessel Pedestal 5503 Drywell Metal Shell Foundation 5334 Reactor Building Basemat 5195 Suppression Chamber including supports 5196 Plant Control Center 5127 Biological Shield 5068 Fuel Pool Slabs and Walls 5039 Drywell Metal Shell 50210 Snubbers 49111 Sacrificial Shield Wall 48912 Reactor Recirculation Piping 48113 Control Rod Driver 47014 Reactor Vessel Core Shroud 46415 Reactor Vessel Nozzle Safe Ends 46416 Turbine Pedestal 46117 Reactor Building Floor, Slabs and Walls 46118 Reactor Vessel Core Support Plate 46019 ECCS Piping Inside Contaiment 45320 Reactor Vessel Core Top and Bottom Grid 45021 Suppression Chamber Vent Headers and Downcomers 44222 Drywell Vent Lines Including Bellows 44223 Reactor Vessel Jet Pumps 44024 Main Steam and Isolation Condenser Piping inside Contaiment 43825 HPCI Turbine 43326 Emergency Diesel Generator (Engine) 42627 Equiment Fundation 42428 CRD Insert and Withdraw Lines 42329 Intake Structure 41830 Cables in Primary Containment 41531 Primary Containment Mechanical Penetration Assemblies 40832 Feedwater Piping inside Primary Containment 40733 Reactor Vessel Feedwater and Core Spray Sparger 39934 Drywell Radial Steel 39435 Main Turbine 39036 LPCI Heat Exchangers 38537 Recirculation Pumps 38038 Main Steam Isolation Valves 37839 ECCS Piping Inside Secondary Contaiment 37640 Refuelling and Drywells Bellows 37341 Emergency Diesel Generator (Generator) 36642 Radwaate Building 36643 Main Condenser 36044 Isolation Condenser Piping Inside Secondary Containment 359

214

45 Essential Switchgear 35846 Cables outside Primary Containment 35747 Turbine Building 35748 Primary Containment Electrical Penetration Assemblies 35549 Essential Relays 34650 Interrptible Bus to Uninterruptible Bus Switchgear 34651 Core Spray Pumps 34352 LPCI Pumps 34353 LPCI Service Water Pumps 34354 Primary Containment Personnel Airlock and Equipment

Hatches342

55 HPCI Pump and Booster Pump 34256 Main Steam Safety and Relief Valves 34157 Isolation Condenser 33858 Reactor Building Structure Steel 33559 LPCI Service Water Piping 33360 Discharge Structure and Canal 33261 Primary Containment Isolation Valves 33162 Fuel Pool Liner 32863 Feedwater Pumps 32164 Offgas Stack 31865 RBCCW Pumps 31466 Moisture Separators 30967 RPS Motor-Generator Sets 30868 Reactor Recirculation Suction and Discharge Valves 30769 Neutron Flux Detectors 30670 Electrical Buses 30671 Bateries 30572 Moter Control Centers 30173 Reactor Building Crane 29874 Instrument Air Compressors 29575 Reactor Vessel Support 29376 RWCU Piping 29377 CS Pumps Motors 29278 Control Rod Driver Hydrulic Unit 29179 Control Room Panels 29080 Reactor Vessel Steam Separator and Dryer Assemblies 29081 SLC Pumps 28882 Circulating Water Piping 28783 LPCI Pumps Motors 28684 Underground Piping 28585 Large Check Valves 28486 Radation Monitors 28387 Turbine Building Structure Steel 28388 Motor Operated Valves 28189 HVAC Ducts 28090 Fire Protection Pumps 28091 Turbine Control and Stop Valves 27492 Gas-oil Transfering Pump 27093 Special and Instruments Cables 270

215

94 Instrument Cabinet (electronic cards) 27095 Low Pressure and High Pressure Heaters 26996 Main Generator 26797 Shut Down Cooling Pumps 26798 LPCI Service Water Pumps Motors 26799 RWCU Regenerative Heat Exchanger and No-Regenerative Hx 263100 Transformers 263101 Drywell Cooling Water Pump 260102 Recirculation Pumps Motors 256103 Recirculation Pumps Motor-Generator Sets 256104 Bateries Chargers 253105 RBCCW Heat Exchangers 251106 Primary Containment HVAC Equipment 244107 Turbine Building Crane 242108 Transmisor Elements 242109 Emergency Filtering Trains 239110 Condesate Tank 236111 Shut Down Cooling Water Heat Exchangers 236112 Temperature Elements 230113 Plant Computer 230114 Essential Local Pannels 226115 Feedwater Pumps Motors 226116 DWR Floor Drain Collector,Sample and Chemical Waste

Tanks224

117 Offgas Recombiner 224118 Oxygem Sampling Pump 224119 Condensate and RWCU Desmineralizers 221120 Gaseous Radwate Dumpers 220121 CRW Floor Waste Collector, Surge and Sample Tanks 220122 Reactor Building to Turbine Building Isolation Seal 219123 Essential HVAC Systems Dumpers 219124 Limit Switches 216125 Equipment Drains Collector Tank 214126 Fire Protection Dumpers 204127 Circulating Water Pumps 204128 Turbine Controls 201129 Cathodic Protection 201130 Shut Down Cooling Pumps Motors 197131 Lube-oil Reservoir and Storage Tank 182132 Fire Protection Pumps Motors 176133 Fire Protection Pumps Diesel Engine 175134 Circulating Water Pumps Motors 159135 Fire protection Instrumentation 157

216

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�� *(1(5$/The following structure is retained for Belgium Equipment Ageing Summaries (EAS):

(TXLSPHQW�$JHLQJ�6XPPDULHV($6 6\VWHPV��6WUXFWXUHV��DQG�&RPSRQHQWV 3UREOHP�FDWHJRU\011 Reactor pressure Core belt region Irradiation embrittlement012 vessel Cover head penetrations Primary water stress corrosion cracking013 Bottom mounted Primary water stress corrosion cracking

instrumentationpenetrations

014 Canopy seal welds Stress corrosion cracking023 Vessel internals Wear021 Baffle bolts Irradiation assisted stress corrosion

cracking022 Guide tube split pins Stress corrosion cracking of Inconel 750031 Control rods Flow induced fretting wear041 Steam generator Corrosion, wear, vibration and fatigue051 Reactor cooling Cast stainless steel casing Thermal ageing052 Pump Thermal barrier Thermal fatigue062 Pressuriser Surge line nozzle Cracking (Fatigue, Primary Water SCC)081 ASME Section III, All Class 1 components Stress and fatigue

Div. 1091 Piping Stress and fatigue due to regulatory

reconciliation092 Stress and fatigue due to unexpected

transients093 Flow accelerated corrosion094 Micro biological corrosion071 Primary piping Cast stainless steel elbows Thermal ageing201 Primary Pre-stressed cables Loss of prestressing

containment202 Concrete Seismical category 1 Leak tightness

structures203 Other structures Degradations due to concrete

carbonation401 IE electrical and 1 E qualification programs and preventive

I&C equipment maintenance402 Electrical cables Cable ageing

Examples of these EAS are given hereafter.

217

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Low alloy carbon steel embrittlement in the beltline region of the vessel. The fracture toughnesscurves (KIc: initiation, Kla: arrest) as a function of temperature are modified as follows:

• shift to the right or increase in transition temperature: RTNDT, corresponding to a decrease offracture toughness at a given temperature.

• decrease of upper shelf values in the transition range.

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The surveillance program, is defined according to 10 CFR 50 Appendix H.

Test specimens representative of the beltline region (shell, weld, heat affected zone) areencapsulated and placed in the RPV. Surveillance capsules are withdrawn according to a predefined schedule in order to, have at all times experimental results for a fluence larger than theactual fluence received by the RPV wall.

The embrittlement is monitored by Charpy impact tests and the shift of the fracture toughnesscurves is supposed to be identical to the shift of the Charpy curves measured at the conventional 41J level.

Fluence is monitored by dosimeters contained in the capsules.

Irradiation temperature is monitored through low melting point eutectic alloys.

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Instrumented Charpy tests: Fracture toughness may be derived from. instrumented Charpy tests bymeans of a correlation between the NDT and the arrest force measured on the Charpy Load trace(force versus displacement). This is a new development not yet accepted in a regulatory context.

Although test methods to evaluate the fracture toughness by means of small specimens (Charpy-sizeor smaller) are under active development world-wide; their use in a regulatory context is notaccepted yet. An increasing use of such methods is however to be expected in the coming years.

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Since the welds of Doel 1&2 are sensitive to thermal ageing and due to uncertainty on Cu content ofthe welds; an enhanced surveillance program is used. This is a combination of the followingtechniques:

• Charpy specimen reconstitution (from broken Charpy remnants);

• Use of load curves from instrumented Charpy tests;

• Determination of fracture toughness on pre-cracked Charpy specimens tested in three pointbending;

• Micro mechanical modelling and micro structural investigations (to understand the factorsinfluencing the RPV mechanical properties under service condition).

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218

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Formulae used to predict the shift of transition temperature RTNDT as a function of variousparameters are:

• The US-NRC Reg. Guide 1.99, rev. 2 formula (not appropriate for low Cu), used for Doel 1/2; -

• The French FIS and FIM formula, used for all units, except Doel 1/2.

Surveillance results are used to confirm the conservatism of these formulae or identify deviationsfrom the predictions.

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Limits on operation: Pressure-Temperature limits for heat-up and cool down and low temperatureover pressure protection defined on the basis of the reference 1/4 T flaw. Stress intensity factor attip of reference flaw must remain < KIa at the given equivalent temperature T-RTNDT. (Criterion:2KIp + KIt < KIa).

Pressurised thermal shock (10 CFR 50 § 50.61): The risk of vessel failure of pressurised thermalshock remains acceptable as long as the following criteria are met: RTNDT < 132C for forging,plates, axial welds and RTNDT < 149C for circumferential welds.

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Reduction of the neutron flux by using core loading pattern of the low leakage type.

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Heating of safety injection water to reduce thermal shock (implemented in Doel 1 &2).

Modification of operating procedure to prevent excessive pressurisation.

�� ($6� �� 5($&725� 35(6685(� 9(66(/� �� &29(5� +($'3(1(75$7,216��35,0$5<�:$7(5�675(66�&25526,21�&5$&.,1*�


Cracks in the penetrations made of Inconel 600.


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ASME VT2 inspections at each outage.

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Detailed inspections of all penetrations in all Belgian plants with Eddy Current Technique; andUltrasonic inspection in case of indications. All heads have been at least inspected once and a newinspection program. was proposed: periodical inspections every 2, 4 or 8 years depending on theunit (if no indications) or more frequently if indications are detected.



The pressure boundary must be still present: minimal ligament length of about 6 mm (4 mm forminimal thickness + 2 mm for the tolerance of the inspection technique) at the 'triple point'


Replacement of the cover at one unit (same design with Inconel 690 material).

219

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The baffle bolts are used to attach the baffle plates to the former plates in the Vessel Internals. Theresulting structure forms a boundary for the flow of coolant and provides lateral support to the fuelassemblies.

After an operating time of the order of 120 000 hours, some bolts exhibit cracking at the junction ofthe head and the shaft of the bolt. This cracking is attributed to a form of Irradiation Assisted StressCorrosion Cracking (IASCC).


Ultrasonic inspections were performed on bolt head of some critical Belgian NPPs (type CPO).

An inspection program should be set up in the future.



Nowadays, cracked bolts are not allowed.


Replacement of all cracked baffle bolts in one unit by new generation of bolts.

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�� ($6��67($0�*(1(5$725��&25526,21��:($5��9,%5$7,21�$1'�)$7,*8(�


Many degradation phenomena that may affect the safe operation of SG’s have occurred world-wide.The following list includes those that have been detected or might potentially occur in the Belgianunits: PWSCC of tubes (primary water stress corrosion cracking), IGA/SSCC of tubes(intergranular attack/secondary side stress corrosion cracking), Tube wear Tube fatigue/vibrationTSP cracking and erosion - corrosion, Wrapper cracking and wrapper support block degradations

Most degradation phenomena are generic in nature and are known to affect many steam generatorsworld-wide, especially those equipped with Inconel 600 tubes.


Tube inspection by eddy current (bobbin coil or rotating tube) and/or ultrasonic examination.Presently by sampling at each refuelling outage.

220



• Percentage of plugged tubes (maximum value per SG, average value per unit)

• Plugging/repair criteria according to structural integrity (crack length, defect depth or signalamplitude), leakage (maximum allowable leakage per SG in normal operation, maximumallowable radiation dose at the site limit in case of secondary pipe break).


Temperature reduction, Chemical cleaning, Chemistry improvements


Tube plugging, Tube sleeving, Nickel plating of the tubes, Tube reexpansion, Tube shot orrotopeening U-bend thermal treatment.

SG replacement.

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The ferrite (10 to 25%) phase of cast duplex stainless steel (austenitic-ferritic) is susceptible tohardening and decrease of ductility, impact strength and fracture toughness under long termexposure to operating temperature.


ASME Code Case N-481 ("Alternate Examination Requirements for Cast Austenitic PumpCasings" - Section XI, Division 1 - March 5, 1990) was introduced to allow the replacement ofvolumetric examination on pump casing welds by visual examination. To allow the elimination ofvolumetric examinations, this Code Case requires:

• an evaluation of the material properties (including toughness) in aged condition, taking intoaccount the degradation by thermal ageing;

• the demonstration of the stability of one-quarter thickness reference flaws, with a length sixtimes the depth postulated at limiting locations.

Re-evaluation of the prediction method during the safety review (every 10 years).


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Mechanical properties of aged material necessary for the fracture mechanics evaluations areindirectly estimated; these evaluations are based on:

• the chemical composition and the ferrite content of the materials as found in material testcertificates;

• the use of lower bound experimental correlation curves provided in the specialised literature;

• the use of property values at saturation.


221

There are no specific regulatory criteria regarding the thermal ageing of cast stainless steelproducts.

The acceptability of the mechanical properties in aged condition was justified by comparing withthe properties given in a generic Westinghouse report (WCAP -13045) applicable to similar NPPs.

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French pump manufacturer Jeumont Industries has observed cracks at three different locations onthe 93 D type RCP thermal barrier and one location in the shaft, under the thermal sleeve, at theinterface of the pump shaft with the thermal barrier labyrinth.

The thermal barrier assembly is intended to provide a thermal buffer between the pumps radialbearing / shaft seal system. and the hot primary coolant pumped, in order to limit the bearing andseals temperature. It is constituted by a stainless steel confinement welded to the pump, mainflange, encasing a heat exchanger fed by the intermediate cooling system (component cooling).

The thermal sleeve is intended to protect the pump, shaft in an area where great temperaturediscontinuity exists.


In service monitoring installed to control. shaft displacements in one NPP (not relevant for all typeof cracks).

No surveillance program related to these specific problems. In Belgium, no frequency defined sofar.



No criteria developed on a generic base for the 93 D type pumps.

However, a specific approach has been developed in 1999 for Doel 3 to justify the continuousoperation of the pumps in service at that time, after discovery cracks during the 1999 UT inspectionperformed on the thermal barrier cover of the spare pump hydraulic, which was removed from aloop in 1993.

The analysis evaluated the crack propagation and the maximal acceptable crack size and showedthat the, number of cycles necessary to reach this depth starting from the detected crack size wassufficiently high to allow a further operation for several years, allowing the implementation of anoptimised replacement program.


New thermal barrier cover design developed in order to remove the geometrical discontinuitycausing crack initiation.

Repair by welding for defect in the flange.

Repair by machining for cracks in the shaft under the thermal sleeve, or replacement in case ofexcessive machining.

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222


Cast duplex stainless steels contain typically 10 to 25% ferrite in an austenitic matrix. These twophases have different chemical composition. The ferrite phase is susceptible to hardening anddecrease of ductility, impact strength and fracture toughness under long term exposure to operatingtemperature.

Aged ferrite is susceptible to cleavage fracture, while austenite is unaffected by thermal ageing. Thefailure mode of aged duplex stainless steels remains ductile tearing (no brittle fracture), but withreduced tearing resistance.


3UHGLFWLRQ: Mechanical properties of aged material necessary for the fracture mechanics evaluationsare indirectly estimated; these evaluations are based on:

• the chemical composition and the ferrite content of the materials as found in material testcertificates;

• the use of lower bound experimental correlation curves provided in the specialised literature;

• the use of property values at saturation.


There are no specific regulatory criteria regarding the thermal ageing of cast stainless steelproducts.

Thermal ageing of duplex stainless steel elbows is deemed acceptable if a leak before break (LBB)analysis performed on these elbows, using properties of aged material, proves successful. In LBBanalysis, the stability of large through wall cracks under accident loading is demonstrated. Thecracks are located in the circumferential and longitudinal welds of the elbows. The analysis isperformed with weld and base material properties.

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The cast stainless steel components present in the primary loop piping of the Belgian units haverelatively low or moderate ferrite content; in this respect, the situation is more favourable than forsome French units where some sensitive components had to be replaced. Thus no replacement isexpected to be necessary for the Belgian units.

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The ASME Section III, Division 1 Code requires that all Class 1 components be qualified for stressand fatigue when subjected to operational pressure and thermal transients classified as service levelsA (normal), B (upset) and test loading conditions.

At the design stage, this qualification is performed using so-called "design transients"; thesetransients and their corresponding number of occurrences are specified in the Safety AnalysisReport (SAR) for each component. However, a re-qualification of some of these components maybe necessary for several reasons.


During operation, transient book keeping must be performed. This monitoring uses the pressure andtemperature sensors existing in the plant. Plant operators record each operational transient having a

223

correspondent design transient as well as all similar variations of operational parameters. Thresholdvalues are defined for those parameters, under these the variations are not recorded. The process ispurely manual and the way to affect operational transients to closest design transients is purelycause-based.

For the oldest units, for which very limited number of transients were considered, the transient listhas been extended to those events considered in the other units. And, for transients not consideredduring the initial period, number of occurrences have been assumed by extrapolation.


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Stress and fatigue analysis according to ASME 111.


At the design stage: The cumulative fatigue usage factor for all transient combinations must be ≤ 1.

During operation: The number of transients experienced by the components cannot exceed thenumber of occurrences assumed in the qualification analysis.


For most units, design transients have been completed and modified for SG replacement and/orpower uprating. Large primary components have been requalified with the new design transients fora 40 years lifetime (if this was not considered in the original design (e.g. Tihange 1)).

Original design calculation are generally very conservative; the main objective of the designer wasnot to, minimise the cumulative fatigue usage factor U, but to get one less than the unity. Therefore,in many cases where U exceeds unity, the most efficient remedy may be to redo the stress andfatigue analysis of the concerned area, using less conservative assumptions and more modemtechniques than those used in the original design.

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According to the ASME Section III code, Class 1 components must be qualified to sustain fatiguedue to design thermal and pressure transients. The qualification analyses are made for a design lifeof 30 or 40 years. Operation beyond the design life must be justified.

In old vintage plants, certain Class 1 pipes and equipment were not explicitly qualified to fatigue(e.g.: old vintage Class 1 pipes were analysed to the ANSI B3 1.1 Code which does not requireexplicit fatigue qualification). In those situations some kind of reconciliation must be performed, letalone when extension beyond the original design life is envisioned.

224


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The equivalence of the design requirement of the ASME III - 68, B31.7 and B31.1 codes with theASME 11183 Code has been established at the first decennial revision of old vintage plants. Thisdemonstration allows to function to 30 years without reanalysing those piping and equipment notcovered by explicit fatigue analysis in the original design.

At the third decennial revision (if any) one subject item should be devoted to:

• the review of the reasoning which led to the demonstration of equivalence between the originalconstruction codes and the ASME 11183 code, to check if it applies to the extended lifetime;

• the evaluation of the significance of the requirements of the new ASME III edition applicable atthe time of the Y decennial revision with respect to those of the 83 edition.


ASME NB-3600 type of stress and fatigue analysis (NB-3200 procedure may also apply when thoseof NB-3600 are not sufficient).


See EAS 081.

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Nuclear plants throughout the world have reported incidents involving cracks, leaks or abnormallarge deformations in some piping systems. After analysis, some of those phenomena wereattributed to thermal transients unknown at the time of the Plant Design. The integrity of theconcerned piping systems and piping supports to these unexpected thermal transients must beverified.


The following transients were observed at least once in at least one Belgian unit, i.e.:

• Thermal stratification transients in the pressuriser surge line, in the main Feed Water lines whenthe Auxiliary Feed Water (AFW) lines are directly connected to them.

• Thermal transients in unisolable section of emergency core cooling system (ECCS) piping thatis connected to the RCS.

Transient monitoring through long term temperature measurement directly onto the concerned pipeshas been performed in some Belgian NPP (e.g. Main Feed Water Lines, Surge Line). This type ofsurveillance allows to determine the time history characteristics of the transients and their numberof occurrences. Transient monitoring must be continuous over at least one fuel cycle because someof these transients only show up for some specific combination of plant operation.

Local fatigue monitoring was installed to control the fatigue usage factor at the point where thestress and fatigue analysis was not satisfactory (e.g. welds between SG nozzles and MFW pipes).

Inspections according to ASME XI.


225

3UHGLFWLRQ: Based on long term temperature measurements, stress and fatigue analysis wereperformed for most of Belgian NPPs.


The acceptability of unexpected thermal transients must be checked through a stress and fatigueanalysis made in accordance with the rules of the ASME Code, Section III, Division 1:

• Subsection NB (NB 3200 or 3600), for Class 1 piping components;• Subsection NC, for Class 2 piping components.

It should however be recognised that the rules contained in Subsection NC to evaluate the fatiguestrength of Class 2 piping components are crude and cannot take into account the local effects ofunexpected thermal transients. Therefore, in severe situations, the rules of Subsection NB may alsobe applied to those components.

The acceptability of cracks must be evaluated with the rules of the ASME Code, Section XI, usingfracture mechanics evaluation technique.


System modifications to avoid or mitigate unexpected transients were implemented in some BelgianNPP:

• a system was devised to depressurise the ECCS portion comprised between leaky ECSS blockvalve and ECSS check valve near the RCS.

• anti-waterhammer baffle boxes have been placed on the main feed water lines near the SGnozzles and avoid the propagation of direct flow lower stratification (which occurs when thesystem is reinitiated after a reactor emergency shutdown.


Repairs and replacements have been applied to Main Feed Water piping components subject toupper stratification (which occurs when low flow of Auxiliary Feed Water is injected into the SG atplant hot standby condition).

Cracked piping components (e.g.: elbow) were replaced in unisolable piping section located nearthe RCS.

Replacement with redesign has been used to MFW piping components such as nozzle to achievesmoother geometry’s with reduced stress raisers (during the Steam Generators Replacement).

Based on problems observed in foreign NPPs, the inspection program may be adapted to inspect assoon as possible critical zones.

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Increasing of the rate of corrosion or material dissolution, caused by relative movements betweencorrosive fluid and a material surface. This is encountered in both single and two-phase flow. Itleads to piping wall thinning.

Flow Accelerated Corrosion (FAC) occurs in many piping systems, mainly FW (Feed WaterSystem), AFW (Auxiliary Feed Water System), BD (Blow-Down System), most of the lines goingto condenser and surge lines. A lot of damaged items have been detected and replaced.


226

Wall thickness measurements by Ultrasonic Testing (UT) of the most susceptible elements areregularly inspected. Inspection campaigns are planned every years with a variable examinationfrequency depending on the severity of FAC of the concerned elements.


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Use of computer codes (checworks) for FAC surveillance.


The criteria for evaluating piping wall thinning are derived form the ASME code requirements.They permit to accept "as is" and for a minimum period of time, wall thinning produced by erosionor by grinding to remove cracks observed during In Service Inspection. These criteria are: theaverage thickness of wall thinning in axial and hoop directions and the extent of wall thinning.


Modification of water chemistry.


Replacements of thinned pipe portions. Sometimes this replacement was performed using anotherpipe material (more resistant to FAC). In few cases: new design of the line.

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The function of these tendons is to sustain the load due to the internal pressure in case of accident inthe containment (design pressure of about 3.5 bar). If the prestress level decreases too quickly, extralifetime of the containment will not be allowed.

This problem is non-relevant for the Doel 1 and Doel 2 (steel containment).


Evolution of the initial pre-stress applied at the primary structures on basis of a minimal period of30 years. A follow-up is defined in the Regulatory Guide 1.90 and in the Safety Report, containingthe following requirements:

• Strain and deformation measurements of the concrete walls by mean of strain cells located inthe concrete or with vibrating wire gauges (permanent monitoring system). These measurementsare performed every 6 months.

• Tension tests for witness cables (lift-off tests). These tests are performed every 5 years.

• Inspection of the protection parts of the anchors of the prestressed cables. These inspections areperformed every 5 years.

• Containment pressurisation tests. These tests are performed every 10 years.

227


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General trend of the results of the lift-off tests.


Acceptance criteria are those from the ASME III Division II CC 3000, the Regulatory Guide 1.90,and the Safety Report.


In some Belgian units, the monitoring system by mean of strain cells was replaced by a monitoringwith vibrating wire gauges.

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Identification of defects concerning leak tightness due to ageing or absence of lining and loss ofresistance of structures by corrosion of reinforcing bars under action of water.

Protection of the risks related to settlements and relative movements between structures.

Risks related to the integrity of the structures due to insufficient reinforcement cover or cracks inreinforced concrete; potentially leak of tightness or confinement due to the joint degradations.


The surveillance is included in "In Service Inspection" program, it consists on:

• absolute and differential settlement measurements of the structures and their foundations(fissurometers for relative movements, topografical levelling for general settlement); and

• visual inspection of structures submitted to thermal and dynamic effects.

The frequency of these inspections depends on the structural importance of the inspected zones(control every year of every 3 years).

If necessary, samples are taken to perform mechanical and chemical tests (e.g. pH measurement).



Minimal mechanical properties of the concrete.

Design settlement values.

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Repair of concrete and injections.

228

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It is imperative to know the qualified life of each equipment or its components to determine the dateof replacement. The qualified life of a 1E equipment is the period during which it is installed andcapable of functioning correctly even in the case of an accident.


1E qualified equipment refers to any electrical or I&C equipment that must be qualified accordingto the regulations in force for Belgium’s nuclear power stations (e.g. the IOCFR50 and a number ofRGs and IEEE regulations) so that this equipment may perform its safety function within the systemof which it is part. An electrical or I&C equipment is considered 1E qualified when it is of identicalmanufacture to an equipment which has successfully undergone the qualification tests.


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Ageing tests simulating the three significant parameters (temperature, mechanical ageing,irradiation) allow to set a qualified life.


It is imperative to replace equipment or their components before their Qualified Life expires.


Qualification programs permit to create data base which will support Preventive Maintenanceprograms by allowing to plan replacements at the time best suited to minimise the economic impact.When a qualification program has been implemented by the relevant tests having been performed inan approved laboratory, at the manufacturer’s or perhaps even at the utility’s premises, the resultswill be recorded in a Qualification Report established by the entity that was responsible forperforming the tests. Based on this report, a Synthetic Qualification Report will be drawn up. Thisreport will contain an evaluation of the qualified equipment life, which may vary depending on thelocation of the equipment.

Extension of Qualified Life: It is interesting to check whether there is a possibility of extending thelife of an equipment by renewing some of its components or by reassessing its life in the light of theseverity of the environment the equipment was exposed to in reality. This approach makes itpossible to have to replace only the equipment exposed to the severest conditions.

Qualification of new equipment: In the case of obsolescence, the Utilities do not have the possibilityto replace different components before the qualified life of the equipment expires. Therefore whenqualified equipment has become obsolete, new equipment has to be found which is alreadyqualified or which we can qualify or have qualified by others under our supervision. Thequalification process may be extremely protracted depending on the required qualification level.

229

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The cabling of equipment and motors, including that of safety-related equipment and motors, of theoldest units, was done with the PVC cables that were available at the time (1970), withoutqualification and without projected lifetime.

The risk exists that a non-qualified cable cannot yet be replaced, simply because an adequatesubstitution for it is not yet available in Belgium. For lack of a qualified supply cable the equipmentitself loses its qualification as it may not be able to correctly operate in the accidental conditions itwas itself qualified for.


1E qualified cables refer to any electrical or I&C cables that must be qualified according to theregulations in force for Belgium’s nuclear power stations (e.g. the 10CFR50 and a number of RGsand IEEE regulations) so that the cable may perform its safety function within the system of whichit is part. An electrical or I&C cable is considered 1E qualified when it is of identical manufactureto a cable which has successfully undergone the qualification tests.

For the original non-qualified cables of the oldest units, the evolution of the mechanical andelectrical characteristics is monitored.


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Ageing tests simulating the three parameters (thermal ageing, moisture ageing, radiation exposureageing) allow to set a qualified life.


It is imperative to replace cables before their qualified life expires.


Qualification programs permit to create data base which will support Preventive Maintenanceprograms by allowing to plan replacements at the time best suited to minimise the economic impact.When a qualification program has been implemented by the relevant tests having been performed inan approved laboratory, at the manufacturer’s or perhaps even at the utilities premises, the resultswill be recorded in a Qualification Report established by the entity that was responsible forperforming the tests. Based on this report, a Synthetic Qualification Report will be drawn up.

This report will contain an evaluation of the cable life, which may vary depending on the location ofthe equipment. These reports are regularly reviewed to assess whether the cable life can be extendedwhen taking into account the exact location and the functionality, new parameters, etc. The reportsare annually distributed to the utilities.

Extension of Qualified Life: It is interesting to check whether there is a possibility of extending thelife of a cable by reassessing its life in the light of the severity of the environment the cable wasexposed to in reality. This approach makes it possible to have to replace only the cable exposed tothe severest conditions.

Qualification of new cables: When qualified cable has become obsolete, new cable has to, be foundwhich is already qualified or which we can qualify or have qualified by others under oursupervision.

230

Cable replacement is in progress in the oldest Belgian NPPs. Cable selection is based according tothe precise location and the function of the equipment supplied through the cables. It follows fromthese criteria that not all the cables have to be systematically replaced, and, therefore, replacementis evaluated on a case per case basis.

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In the component service life assessment approach, the vessel is dealt with separately, owing to itsspecial safety nature which is linked to its role as a secondary barrier, which leads to its rupturebeing excluded and, besides, to the fact that its replacement is very difficult to conceive.

Regarding the vessel service life, the main ageing factor concerns the embrittlement under neutronirradiation of the core region. The difference in ductile-brittle transition temperature which resultsfrom it is assessed, using prediction formulae derived from tests in the research reactor andfeedback of experience. Embrittlement has been taken into account from the design stage of the firstFrench PWR units, by preventive measures aiming to reduce the content of embrittling elements,such as copper, phosphorous and nickel.

The predictions are then checked for each vessel with the aid of the monitoring programme, basedon the use of test specimens of material representative of the shell and welded joints; these areirradiated at the edges of the core, so that the real evolution of the transition temperature can beanticipated.

The extrapolation of the results, combined with all the studies and R&D work conducted up tillnow, confirms the in-service behaviour of 900 MWe unit reactor vessels for a service life of at least40 years, without implementing restrictive protective measures, but requiring additional actions(optimised in-service inspection, treatment of irradiation surveillance programme anomalies, 3Dmechanical thermohydraulic calculations, design of certain state-oriented approach procedures etc.).

The scientific work conducted, notably concerning ageing mechanisms by irradiation of the vessels,has resulted in a large acquisition of knowledge on the evolution of the transition temperature inview of irradiation (RTNDT) until the end of service life of the reactors, linked to the fluencereceived by the vessel in the core region. The method used to assess fluence is today qualified, as aresult, in particular, of the special monitoring programme and the autodosimetry of the CHOOZ Avessel.

Moreover, fuel loading patterns at reduced flux, making it possible to limit vessel embrittlement,are implemented on most of the 900 MWe CP1-CP2 units and extended to CP0 units. Thisoptimised management of fluence is also being developed on units 1300 MWe. It will allow, inaddition to limitation of vessel embrittlernent to compensate, the flux effects linked, to long series.

A summarised document confirming the inservice behaviour of REP 900 MWe vessels for et least40 years, was drawn up and submitted to the safety authorities at the beginning of 1998 (1). thesame will be done for the REP 1300 MWe vessel file.

The briefing book issued includes, in particular, the following points:

• The "available margins" have been updated from the actual fluence level, with the computingsequence which today constitutes the reference for flux calculations, and the calculated fluencevalues of each vessel are lower than the design value, owing to setting up of optimisedmanagement. The "available" margins are increased.

• The thorough knowledge of the evolution of RTNDT up to the end of reactor service life linkedto the fluence received by the core region. All of the estimated reactor end of life RTNDT ,

231

according to their commission date, show that the values are much lower than 1000°C,especially in view of the saving of fluence achieved by optimised fuel management.

• The margins with regard to fast fracture obtained in mechanical analysis, demonstrate theacceptability, in relation to code criteria (RCC-M, RSEM), of assumed underclad cracking, aswell as the harmlessness of assumed faults in the cladding for all loading situations.

• The confirmation, by specific studies, of rare indications detected on some vessels, has beengiven for the different loading situations.

A reactor service aptitude criteria has been suggested, in line with the approaches and criteriainternationally selected in the assessment of reactor integrity (USA - screening criteria, Germanyetc.) and is based on the results obtained within the scope of the briefing book and additionalstudies. This reactor service aptitude criterion, indexed in terms of measurable parameter RTNDT,is intended to look at the acceptability of a 130°C value.

The methods selected for the in service inspection applied for the examination of vessel walls andthe internal condition of the core shells, have been used with the checking device termed "ThirtyFirst Millimetres" fitted onto die in-service inspection machine. The aim is to guarantee thedetection and characterisation of indications of height greater than 6mm under cladding whateverthe surface condition of the latter (the vessels shall be examined during the service visit after 20years of operation at the latest).

Given what is known about the resilience of the materials following the studies relating toembrittlement under irradiation, the good assessment of fluence at the end of service life andmargins identified in the mechanical analysis for the most severe loading situations, a service lifebeyond 40 years is conceivable. Therefore, an extension to the present vessel monitoringprogramme has been decided. It consists of inserting, in place of irradiation capsules which havebeen removed, reserve capsules, in order, in respecting regulatory requirements, to supply thenecessary factors for a possible service life extension, and to reinforce the embrittlementassessments made and to reduce uncertainties.

The comparison of the French approach to embrittlement by irradiation with that of those practisedabroad, has made it possible to foresee that the vessels in our inventory possess a real potential interms of service life, owing, in particular, to the materials used, the quality of the monitoringprogramme and the fluence management policy implemented. EDFs participation in differentinternational programmes currently underway confirms this analysis.

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Some parts of the main primary system, In particular the bends, are in moulded cast stainless steel.Those components are created in a foundry, by a static casting or centrifugation process, resultingfrom technical and economic considerations in the 1970's.

In the early 1990's, it was noticed that those moulded products could present thermal ageing, afterbeing maintained for a long time at main primary system service temperature (hot leg at 320°C orcold leg at 285°C). This results in a modification of the material structure, and the effect of this is tomake the mechanical properties become hard and embrittled, which is demonstrated by a weakeningof resilience and resistance.

The materials concerned by the phenomenon are: the moulded bends of the hot and cold legs, andthe primary coolant pump casing, the injection line shut-off valves, the inclined tap and pressurizerspray valves.

Large scale work has been undertaken for a number of years with the aim of assessing their servicelife. They can be grouped around two large themes:

232

• The search for ageing mechanisms and the parameters that govern it, with determination offorecasts of the resistance characteristics at rupture at end of life.

• The demonstration of the strength of the parts in service by calculations and tests, completed bya defence in depth following a leak before break approach,

The scientific programme has rapidly progressed since 1993. The ageing mechanism of theseproducts moulded in cast stainless steel has been perfectly identified. Tests on large-scale modelsand test specimens have shown that this steel, even aged, remains ductile, which rules out any riskof brittle fracture.

The formulation of tearing strength at end of life forecasts has been established from the ageingmonitoring programme results. This consists of ageing, in an accelerated way, representativeproducts (cast ingots moulded at the same time as the parts) as well as "casing" products of thecomponents in service, on account of their chemical compositions.

In order to validate these forecasts, a large expert assessment programme has been started. Itincludes, in particular, the removal of sensitive bends during steam generator (SG) replacements.The bends removed in 1995 for the DAMPIERRE unit 3 (3 hot bends and 2 cold adjacent to theSGs) are being expertly assessed, The first results show that the measurements are either confirmed,or better than the forecasts. Five other bends, considered as the most aged, will be expertly assessedduring future SG replacement operations.

The checking by gamma radiography of the bends at input and output of the most sensitive SGrevealed no abnormality questioning the quality of the manufacturing. The non destructive controlmethod R&D work has made it possible to calculate the readings from the radiograms and showedthat they were small.

The mechanical strength of the moulded bends has been demonstrated by a number of calculationsand tests on age bonds at range 1 and 2/3. These tests have led to a very limited opening of faults.

This work, which has no international equivalent has made it possible to confirm the operatingfitness of all the hot and cold bends for the most unfavourable mechanical load situations for 900MWe units, for a service life, of at least 40 years.

In EDFs opinion no bend replacement is therefore necessary for safe operation. Nevertheless,accompanying measures have been taken and begun in order to confirm the acceptable behaviour ofthese components, notably the non-development of their real defects, and in order to assess theseverity of the ageing predictions.

Among these, besides the removing of hot and cold bends for expert assessment during SGreplacement operations, an ageing monitoring programme of the most representative hot bends hasbeen selected, based on the ageing monitoring in a furnace of cast ingots and coupons taken fromthe removed bends. Furthermore, at each decennial outage program. an inspection programme bygamma radiography of some of the most representative SG input and output bends will be carriedout in order to confirm the absence of development of the defects previously detected.

The 900 MWe unit bend justification documents were presented to the safety authorities in 1996and 1997, who noted large-scale work presented by the owner. It reached a favourable conclusionfor cold and hot bends, except for eight of them, considered as being the most sensitive to thermalageing, and for which the maintaining in operation is still dependent on an additional analysis to beprovided in three years time.

With regard to 1300 MWe unit moulded bends, the operating fitness justification for at least 40years has been supplied, and the justification documents have been submitted to the safetyauthorities. This documentation was established using the same methods that were selected for the900 MWe plant: the results are similar. As an accompanying measure, inspection by radiographywas selected for a SG input and output bend at each decennial outage program for one unit. The

233

ageing monitoring program, begun for components of 900 MWe units, has been extended to somecomponents representative of the 1300 MWe plant.

At the end of 1997, a document demonstrating the operating fitness, for at least 40 years, of primarycoolant pump casing, of some tap and valve casing and inclined nozzles of the Safety InjectionSystem (RIS) injection lines in the primary system, was submitted to the safety authorities. Thisdocument, which is the last justification section for the strength of primary system mouldedcomponents service life is currently being examined. Additional requests could be formulated bythe safety authorities on the part of the document relating to inclined nozzles.

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With regard to service life, the problems likely to affect control have very different origins:ageing, obsolescence of components and material, operating developments. Up until now,none of these problems has caused any considerable difficulties, even if some partialreorganisation/recasting operations have been necessary.

The aim of forecasting in this area, in order to control the safety of the unit, is to guaranteethe medium or short term maintenance of the systems, by looking for a technical andeconomic optimum.

The lasting quality problematic of the nuclear unit instrumentation and control systems hasbeen considered, since 1987, within the general framework of the "service life"programme, which defined instrumentation and control as one of the high stakecomponents for unit service life. The inspection report derived from these studiesconsiders the possibility of retaining instrumentation and control equipment for a period of20 to 25 years after the industrial commissioning of the units. To reach this objective, anaction programme and policy, forming the approach termed "lasting quality", have been setup.

In addition to this approach, studies have been conducted in order to identify the differentpossible scenarios for the updating of instrumentation and control systems. The conclusionof this phase, presented in 1993, lead to the following recommendations:

• A limited redevelopment, carried, out in stages and in line with a coherent targetarchitecture diagram.

• The launch of preliminary opportunity and feasibility studies of the "Instrumentation andcontrol redevelopment" of 900 MWe units (R2C project).

The R2C project strategy studies were conducted, from 1993 to 1995, in preparation forthe 900 MMe plant second decennial outage programme.

Using the R2C project results, the 1300 MWe unit study, equipped with digitalinstrumentation and controls more sensitive to obsolescence, was launched at the end of1996. It is subject to the ACCORD project (Instrumentation and control analysis, decennialredevelopment optimisation) for which conclusions have been drawn.

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From the risk analysis, the lasting quality strategy has distinguished two groups ofinstrumentation and control system suppliers:

• The main equipment suppliers for which long term. agreements arc desirable

• Other suppliers for which the use of existing maintenance structures arc desirable.

For each supplier of the first group, the following actions have been taken:

234

• Negotiation of a lasting quality protocol, for a 25 year period, which defines theactivities to be perpetuated.

• The setting up of protocol application contracts, generally renewable every three years,allowing the activities defined in the protocol to be ensured.

For the suppliers in the second group, the strategy consists mainly of forming spare partsupplies, to consider, when necessary, the equipment replacement by functionallyequivalent systems and sometimes to ensure, by contract, maintenance over 5 to 10 yearduration.

Within this framework, some 30 contracts have been negotiated with 900 and 1300 MWeplant instrumentation and control system suppliers. The corresponding strategies arecurrently being defined for the N4 plant equipment.

In practice, the feedback in experience of these lasting quality actions shows that theforecasting carried out has a wider spectrum than. that originally considered. This is thecase, for example, in the field of training of owners in maintenance and management.

In fact, the concentration of skills contract which plans for remote control "technicalsupport", from the designer to the plant operator allows, by analysing either the number ofrequests, or the same content of these requests, to get an accurate idea of the level ofcompetence of the on-site participants.

This analysis, which is carried out on a yearly basis, allows deviations to be demonstratedIt is then possible to alert the management of the plant concerned and thus anticipate,before an operating incident, any possible bringing up to scratch of skills. This loop periodshould expand insofar as the reliability of modem instrumentation and control systems isbetter, by one to two decades, than that of our older systems and that they are in general alot more complex.

Improvement in the reliability and increased complexity means that skills "refresher"courses are necessary, the analysis of technical support is an indicator that allows thistraining to be judiciously anticipated.

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The objective of the R2C project was to establish a master plan for the 900 MWe unitinstrumentation and control for the period of their second decennial overhaul (1998 -2007). The studies concerned the large-scale industrial computing, instrumentation andcontrol, of which the redevelopment can only be contemplated during an outage of longduration.

With regard to the electronic components (mainly analogue on these units), the "age-monitoring body" concluded on the maintenance feasibility of most of the equipment, up tothe date set for the third decennial outage progams.

These conclusions are based on the lasting quality approach which guarantees at the leastthe support of the suppliers during this period. Additions have been made to the lastingquality contracts in order to reinforce the forecast of the treatment of obsolescence of theelectronic components.

The maintenance average annual cost forecast linked to this extension remain a lot lowerthan the equipment updating investment assessments.

However, some updating has been decided on, in order to resolve the ageing problems ofsome sub-assemblies. The cause of some of the ageing observed is excessiveoverheating due either to bad equipment design, or to inadequate air conditioning in the

235

building. According to the additional functional requirements to be taken into, account, theredevelopment strategy has lead to the following:

• For the reactor protection threshold detection systems, to conserve the technology andrebuild functionally identical analog modules.

• For the flux measuring system, to adopt numerical equipment that improves the testand maintenance facilities.

The power component ageing expert assessments (diodes, thyristors) also showed thatsome of these components were nearing their end of life, on account of design orproduction defects. Preventative replacements were organised before these componentsbrought about the failure of the equipment in operation.

Besides, operating modifications have been included in the project conclusions: updatingof the level control of the CP0 unit steam generators and extension of the number ofalarms for all 900 MWe units.

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The objective of the studies forming part of the ACCORD project, in preparation for thesecond decennial outage programmes for the 1300 MWe plant units, is to analyse thelasting quality of the instrumentation and control systems and to propose maintenance orredevelopment strategies for this equipment.

The digital systems, used for the 1300 MWe plant, were designed in the late 1970’s. Theanalysis of the list of components implemented allows a short term obsolescence to beexpected (5 to 10 years) for all these components. The treatment at irregular intervals ofobsolescence, practised up until then within the scope of lasting quality contracts is nolonger adapted to deal with the extent of these obsolescences.

However, the stability context looked for the whole series does not allow the "concentrationof skills" of the lasting quality contracts to be maintained in good conditions, owing to thevery small volume of modifications to be studied. The maintaining in the long term of thesedesign skills, still existing with suppliers, therefore seems very difficult.

These observations encourage the equipment to be fixed and to guarantee their hardwaremaintenance by storage in advance of all the components. This storage, by EDF, of all thecomponents necessary for identical repair of the electronic equipment, until the end of lifeof all units, bas been decided (design, organisation). The first assessments of this strategylead to a purchase cost, validation and management of the stock in the region of 1% of theequipment price.

In order to cover the possible failure of perpetuated suppliers, which today carry out therepairs, an analysis of the transfer conditions of these repair contracts to another serviceprovider is currently underway.

The expert assessment of electronic equipment has underlined the risks linked to the wearof the plug-in type connection. No servicing industrial solution seems available to correctthis wear, The only means of action available are to strictly limit preventive maintenanceoperations and to improve the breakdown diagnostic facilities and procedures, in order tolimit the number of connection/deconnection manoeuvres of the electronic cards.

Providing that the functional requirements are fixed and the obsolescence problems amanticipated by storing components, the maintenance of 1300 MWe unit instrumentationand control equipment must therefore be equipped to be extended at least until the thirddecennial outage program, for an estimated overcharge of about 10% of the averagemaintenance cost observed today. It is not the same for industrial computerised systems,

236

whose functions are still developing in a very dynamic industrial context (computer rate ofobsolescence increasingly rapid).

Strategic redevelopment studies have been launched for the sub-assemblies withindustrial computers or for materials that are not covered by lasting quality contracts.

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In line with actions recommended for the 900 and 1300 MWe plants, an analysis iscurrently underway to anticipate the storage of components necessary and the repair ofinstrumentation and control equipment of the N4 plant.

More generally, consideration has begun to incorporate, in future equipment contracts,clauses to take into account maintenance in operating conditions, from the design of thesystems. This approach requires these aspects to be included in the specifications and tothe offer analysis criteria to be defined. The objective would be to obtain, for consultation,an assessment, on the same basis for all offers, of the ownership costs of the system to besupplied. A particular skill must be developed at EDF in order to cover, in the best possibleconditions, these important stages of the purchasing process.

Considerations and feasibility studies are currently being launched, within the scope ofR&D, in order to assess the methods allowing to maintain the functions of theinstrumentation and control systems by a succession of small changes/developments ofthe equipment, while ensuring the control of downtime and maintenance of skill.

The periodical reassessment of the partial updating and maintenance strategies mustallow to reliably retain the quality of the instrumentation and control functions, withouthaving au impact on the unit service life. Up until now, there seems no need tocontemplate a major revision of the instrumentation and controls of the 900 and 1300MWe units. However, on the third decennial outage program the possibility that moresizeable operations may be required should, not be ruled out.

�� 5($&725�&217$,10(17The pressurised water reactors are inside of a reactor building, which ensures containment, and inthis way protects the public and the environment against radioactive products likely to be scatteredinside the enclosure in an accident situation.

To this end, the reactor containments have been designed to withstand accidents leading tomaximum pressure values and to present acceptable leakage in these conditions.

Thus, with regard to the reactor containments of different plants, two aspects are to be consideredwhen assessing service life : resistance and leaktightness,

The inventory containments are of two types:

• CPY plant containments, made up of a single prestressed concrete wall, lined on its internal facewith a metallic leakproof coating.

• The P4, P’4 and N4 plant containments made up of a double wall : the internal wall inprestressed concrete and the external wall in reinforced concrete. Possible leaks from theinternal wall are collected in the inter-wall area maintained under negative pressure and treatedbefore being discharged into the atmosphere by means of the suction and filtration circuit.

The phenomena likely to influence the service life of the structures are mainly linked to theshrinkage-creep of the prestressed concrete which results in a reduction in tension of theprestressing cables throughout the life of the units.

237

In view of the impossibility of re-intervening on the cables to restress the building as they arecement grouted, in ail areas of the internal wall the residual stress at the end of the theoreticaloperating life of the units should be known in order to ensure that the containments do not exceedthe leaktightness criterion in an accident situation.

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The monitoring of reactor containment behaviour is carried out for leaktightness and the strength ofthe structure:

• During containment pressurisation, by measuring the leak rates and by checking deformationsand displacements under the effect of pressure variations.

• During normal operation, to increase knowledge about the kinetics of the concrete shrinkage-creep phenomena directly influencing the prestressing losses and to make clear the integritystatus of the internal containment around the penetrations.

Mechanical behaviour, both during tests and over time, is monitored with the aid of the testingdevice, made up strain gauges, pIumblines, invar wire and dynamometers, whose distribution on thestructure allows the results obtained from different types of sensors to be compared, and a resultingstatement to be made.

Dynamometers intended to measure prestressing cable tension are fitted on four vertical cables,injected with grease, of the first unit of each site (the "lift-off" technique, implemented in 1996 byway of experiment on the CIVAUX and FLAMANVILLE sites, has made it possible to state thereal tension-of the corresponding cables).

The measuring frequency of the whole of the test device is done quarterly at a minimum. But withthe aim of improving knowledge about the behaviour of some containments, telemetering systemshave been installed.

In addition to the fact that they allow more measurements to be made in normal operatingconditions and during containment tests, during particular episodes in the unit life (reactor outageand start-up, exceptional atmospheric conditions etc.), measurements can be performed morefrequently and in this way knowledge of containment thermal behaviour can be considerablyimproved.

With regard to measuring the leak rate in operation, the SEXTEN system essentially makes itpossible to detect leaktight faults at the level of penetrations. Version 2 makes it possible not only toincrease system availability but also to obtain greater efficiency in the leak search.

238

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CPY plant containments

The behaviour of the 900 MWe unit containments is satisfactory. An exemplary respect for theleaktight criteria is shown, a consequence thought to result from the presence of the metallicleaktight coating.

Results obtained in tests have clearly demonstrated satisfactory mechanical behaviour of thecontainments under the effect of pressure variations. With regard to evolutions over time, a goodoverall homogeneity can be observed on the plants, with acceptable ranges compared with thetension loss in the prestressing cables, and a clear reduction of the phenomenon on most units.Some of them demonstrate lower ranges.

It appears, therefore, that there are no notable evolutions likely to challenge the respect for the 40year criteria. Nevertheless, endeavours will be made to check that the shrinkage-creep phenomenaobserved on these containments have no consequence on the performance of the leaktight coating.

N4 and 1300 MWe plant double-wall containments

The results obtained up till clearly show shrinkage-creep phenomena greater than those taken intoaccount in the design and, consequently, also greater prestressing losses.

The first studies conducted. on the P4 and P14 containments had demonstrated that, the residualcompressive stress would respect for almost all units, at 40 years, the minimum value set by thesafety reports.

However, in recent tests, some containments presented an internal wall leak rate which exceeds thetest criterion of the safety report. On these containments were observed cracking around theequipment access hatch. This cracking is brought about by prestressing losses linked to the largeconcrete shrinkage-creep in an area where the initial prestressing is weaker than in the continuouswall-bead owing to the deviation of the cables around the equipment hatch.

On the other hand two containment have presented since they were constructed, leak rates whichstem from leaks distributed over the whole of the structure.

In view of these recent observations, a containment ok grading has been carried out, which has leadto three structure categories being defined:

½ Confirmed sensitive containments: those which do not respect the test criterion for internal wallleak rate set by the safety report (1% / day of the air mass contained in the containment),

½ Potentially sensitive containments: those which are considered likely to exceed the 1 %/daycriteria in the next test,

½ non-sensitive containments. those which, given either smaller shrinkage-creep ranges or arestrained leak rate, measured in the last test performed, have a low probability of exceedingcriterion in the next test.

However, on the one hand, current studies show the capacity of the Containment AnnulusVentilation System to deal (in case of accident) with sizeable leaks of the internal wall, at least upto 5% / day, and on the other hand, all the measurements made have made it possible to check thatthe leak rate, not passing through the area between the walls, has still remained lower than thethreshold set to respect the direct leak rate (with embedded basemat).

This makes it possible to demonstrate that, despite the performance losses observed, the safetycriteria, notably in terms of radiological consequences, would be respected. Moreover, the nature ofthe cracks, and their behaviour in accident situations, makes it possible to demonstrate that exceptin exceptional cases, the leaktight criteria set in the stress report.

The treatment of the problems identified is subject to:

239

½ Resistance studies of the structures to determine, unit by unit, the tension areas likely to appearbefore the end of the 40 year limit in the sensitive parts of the containment (equipment accesshatch areas, the dome-cylinder valve body crotch, the runners.

½ The definition, for the remaining unit service life, of a servicing method to reduce leaks byapplying a composite clad on the areas where cracks have appeared.

½ The material and functional demonstration of reliability and efficiency of the ContainmentAnnulus Ventilation System.

½ The overall analysis in the long term of the containment of the double-wall reactorcontainments. It better determines the correspondence between the air leaks measured in testsand those in air and steam which occur during an accident in order to confirm the respect of theauthorisation decrees which set the acceptable leak criteria of the internal containment at1.5%/day in case of accident.

½ The start of a R&D programme, running over several years, and intended to:

• On the one hand, better grasp the modelling of the different prestessed concretedeformations under biaxial loading.

• On the other hand, to validate, on the scale of a structure of industrial dimensions, the leaktransposition in air compared with the air and steam leaks, as well as the resistance of acomposite clad in the thermal conditions of a MAEVA model project.

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The ability of reactor containments to fulfil, for 40 years, the functions for which they weredesigned, is globally achieved for 900 MWe plants.

For double-wall containments (P4, P’4 and N4), the results obtained lead one to believe that theleaktightness observed on the internal wall of some units is likely to occur before 40 years (in theequipment hatch area as well as in other areas that are considered sensitive). The ContainmentAnnulus Ventilation System has demonstrated its ability to deal with, in case of accident, internalwall leaks at least up to 5%/day.

The studies begun, since the discovery of certain discrepancies, must, however, make it possible tosuggest lines of defence which will ensure that adequate leaktightness and integrity is maintainedfor structures and that the stress report is respected with the objective of a 40 year service life.Nevertheless, individualised forecasts and follow-ups are necessary.

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Cables were originally considered as "sensitive equipment". This is not due to the direct cost ofrecabling a unit (estimated at 2.5% of the unit price), but to the length of such an operation,estimated a nearly 1.5 year, which would result in a very high indirect cost (without counting all therequalification tests).

240

Two sets of cables can be differentiated by the material making up the casing and the insulation:

• The function of which is required in reactor building during and/or after accidental condition(said K1) cables, insulated with ethylene-propylene rubber (EPR) and cased in chlorosulfonatedpolyethylene (CSPE). There are much smaller in number, with a total length of less than 50kmper unit.

• The other cables, insulated and cased in PVC. They represent most cables, with a total length ofabout 1000 km for a 900 MWe unit.

Since this time, another material has appeared for cables "without halogens", ethylene vinyl acetate(EVA).

Although the feedback of experience for conventional units may lead one to expect a cable servicelife greater than 40 years, some doubts remain:

• Thermal ageing under irradiation of PVC cables.

• Ability of K1 cables to carry out their function during a thermodynamic accident occurring atend of life.

In this context, specific studies have been conducted in order to find out the service life for thesecables in nuclear environments.

The objectives can be summarised in the following way:

• Search for parameters sensitive to ageing of material from which the cables are formed.

• Determination of the prominent degradation mechanisms.

• Determination of long term behaviour in normal conditions, on mechanical operating criteria(by modelling on the strain property at rupture) and on electric operating criteria (insulation,dielectric resistance),

• Determination of long term behaviour in accident conditions, on electric operating criteria(insulation, dielectric resistance).

• Search for non destructive inspection tools for expert assessments.

• Creation of a cable surveillance methodology.

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The characterisation (mechanical/electric/physical and chemical) of cable samples is carried out onparameters sensitive to ageing.

The sensitiveness of numerous parameters has been studied for constituent material. The two mainconclusions are:

• The ageing of the materials studied does not significantly affect the electric criteria linked to thefunctionality of the cable, with the exception of PVC, for which the evolution of insulatingresistance is mainly attributed to the migration of plasticizers, in some formulation patterns(presence of phosphate, of chlorinated paraffin. at the level of the insulating enclosure).

• Strain at rupture is a parameter sensitive to ageing for all materials.

The experimental data, collected in the different test series accelerated under thermal and radiativestress, have made it possible to model the evolution of strain at rupture according to time,temperature and dose rate variables and to 6 parameters characteristic of the material studied. Theknowledge of this mathematical function allows a predominance diagram to be constructed whichprovides direct access, depending on the dose rate and temperature conditions applied, to the type of

241

degradation to which the material is subject. The model has been validated on the four materialsstudied.

We have also shown that this kinetic model, averaging some adaptations, perfectly describes theprediction models developed by S.Burnay and KT.Gillen whose work is referred to internationallyin the field of cable ageing.

Knowledge of the model parameter values allows the material service life in normal operatingconditions to be calculated (temperature and dose rate fixed respectively at 50°C and 0.1 Gy/hr) foran end of life criteria of 50% absolute of the strain property at rupture.

For the four existing materials on the French units, the estimated service life considerably exceeds50 years.

For the Kl cables, which must carry out their function in case of thermodynamic accident, we havestimulated the accident conditions according to the design basis accident (DBA) specification andhave analysed their behaviour. These analyses were done both on new and aged cables, in a naturalway, taken in the plant after 12 years of operation.

For the cables studied in EPR-CSPE and without halogen, the DRA resistances are acceptable oncomplete cables. Thus, the cables studied can withstand an DBA after 50 years under normaloperating conditions.

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The control of cable ageing, mainly those installed in the reactor building, relies on:

• Qualification tests in accident conditions,

• Laboratory tests and studies to establish a prediction model,

• The many expert assessments carried out on suite in response to an incident or during updatingoperations of conventional thermal unit instrumentation and control.

It is on the basis of these results that we have created our cable surveillance methodology anddefined two groups of cables, according to the environmental conditions in which they are used.

The first group is that of cables operating in normal conditions (air temperature lower dm 50"C,dose rate lower than 0.1 Gy/hr, dose accumulated over 40 years less than 35 kGy, chemicalenvironment neutral). In these conditions, the cable service life is at least 50 years,

Monitoring for these cables consists of a routine visual examination of the ends which are the areasmost subject to degradation. This is done during servicing of the sensors or actuators and electricpanels, or the instrumentation and control cabinets or bushing.

The second group is that of cables operating in more severe conditions than normal. In this case,routine monitoring of a representative sample shall be carried out in order to ensure that theiroperating capacity is maintained.

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The ageing models, established from laboratory accelerated ageing tests, the expert assessments, theaccident and ageing tests performed on cables taken from the power station, have confirmed that ,under normal operating conditions, power station cables have a service life greater than 50 years,including K1 cables, which are able to withstand an accident after 50 years of normal nuclearoperating conditions, Studies have been started with the aim of examining the ageing of cables usedon an irregular basis outside the conditions planned in the original specifications.

Consequently, no large-scale replacement of cables is to be considered. for a 50 year period.Nevertheless particular attention should be paid to the identification of the most stressed cableswhich are subject to a harsh environment and to their replaceability en a forecast basis.

242

The possible change of a limited number of cables, further to operation in harsher conditions thannormal, does not appear to pose major technical problems.

A briefing book, stating the strategy for qualification in accident conditions, ageing modelling andmaintenance methodology, is currently being drawn up.

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�� *(1(5$/The following degradation mechanisms have been identified for Spanish Plants:

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SCOPE OF MAINTENANCE PRACTICES EVALUATION

&RPSRQHQW 'HJUDGDWLRQ�0HFKDQLVPVReactor Pressure Vessel Neutron Embrittlement, Fatigue, IGSCC, WearSteam Generators Fretting, SCC, Fatigue, FoulingPressurizer Fatigue, IGSCC, Electrical AgingRPV Internals Wear, IGSCC, IASCCContainment Mechanical PenetrationAssemblies

Galvanic Corrosion

RPV Support Corrosion, Fatigue, SCCEmergency Diesel Generators (Engine) Wear, Fatigue, Corrosion, Stress Relaxation,

Fouling, Erosion, MICMain Turbine Fatigue, Erosion/Corrosion, SCC, WearMain Generator Fatigue, Electrical Aging, WearReactor Coolant Pump Fatigue, Thermal Embrittlement, Acid Boric

Corrosion, SCCFeedwater Pumps FatigueEssential Service Water Pumps Corrosion,Charging Pumps Acid Boric CorrosionFeedwater Turbine Corrosion, SCC, FatigueEssential Service Water Piping Galvanic CorrosionReactor Coolant Piping Thermal Embrittlement, FatigueControl and Volume Piping FatigueResidual Heat Removal Piping FatigueSafety Injection Piping FatigueMain Steam Piping FatigueFeedwater Piping FatigueMain Turbine Piping Fatigue, Erosion/CorrosionCirculating Water Piping Erosion, CorrosionSteam Extraction, Drain and Vent Piping Fatigue, Erosion/CorrosionMain Steam Isolation Valves Erosion/Corrosion, Erosion, Fatigue, WearMain Steam Relief Valves Fatigue, Erosion, CorrosionFeedWater Control Valves Erosion, WearSafety Injection Accumulator Check Valves Acid Boric Corrosion, ErosionMain Steam Safety Valves Fatigue, Corrosion, Erosion, Stress RelaxationComponente Cooling Heat Exchangers Erosion/Corrosion, MICMain Condenser MIC, Fretting, Crevice, Galvanic CorrosionRadwaste Monitoring Tank Pitting

243

Essential Chillers MIC, CreviceMotors Wear, Electrical AgingMCC and Load Centers Wear, Electrical AgingContainment Electrical Penetration Assemblies Electrical AgingTransformers Electrical Aging, FoulingCables Electrical AgingReactor Building Cracking, Armature CorrosionIntake Structure Erosion, Armature CorrosionDischarge Structure Erosion, Armature Corrosion

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SCOPE OF MAINTENANCE PRACTICES EVALUATION

&RPSRQHQW 'HJUDGDWLRQ�0HFKDQLVPVReactor Pressure Vessel Neutron Embrittlement, Fatigue, IGSCCRPV Internals Wear, IGSCC, IASCC, FatigueContainment Mechanical PenetrationAssemblies

Galvanic Corrosion

Metal Containment Corrosion, Galvanic CorrosionDrywell Vent Lines Including Bellows FatigueReactor Building Crane OverloadTurbine Building Crane WearRPV Support FatigueEmergency Diesel Generators (Engine) Wear, Fatigue, Corrosion, Stress Relaxation,

Fouling, Erosion, MICMain Turbine Fatigue, Erosion/Corrosion, SCC, WearMain Generator Fatigue, Electrical Aging, WearRecirculation Pump Fatigue, Thermal Embrittlement, Acid Boric

Corrosion, SCCFeedwater Pumps Fatigue, WearHPCI Turbine SCC, FatigueRBCCW Pumps Wear, FatigueCirculating Water Pumps WearLPCI Pumps WearCore Spray Pumps WearHPCI Booster Pump WearService Water LPCI Pumps WearShut Down Cooling Pumps WearFire Protection Pumps Wear. FoulingECCS (LPCI, HPCI, CS) Piping insideContainment

Fatigue, IGSCC

Main Steam Piping inside Containment FatigueIsolation Condenser Piping Fatigue, IGSCCRecirculation Piping Fatigue, IGSCCUnderground Piping Galvanic Corrosion, MICCRD Insert and Withdraw Lines IGSCCFeedwater Check Valves Erosion/Corrosion, Erosion, Fatigue, WearRecirculation Suction and Discharge Valves Thermal Embrittlement, Fatigue, IGSCCMain Steam Isolation Valves Erosion/Corrosion, Erosion, Fatigue, Wear

244

Main Steam Safety and Relief Valves Fatigue, Erosion, Corrosion, Stress RelaxationRBCCW Heat Exchanger Crevice, MICMain Condenser Galvanic Corrosion,Pitting, Erosion/Corrosion,

CreviceDWR Floor Drain Collector, Sample andChemical Waste Tanks

MIC, Crevice

Motors Wear, Electrical AgingMCC and Load Centers Wear, Electrical AgingContainment Electrical Penetration Assemblies Electrical AgingTransformers Electrical Aging, FoulingCables Electrical Aging

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Irradiation embrittlement is the consequence of the alteration of crystalline metal net due toneutronic effects. The main variables controlling the damage are: Ni content, Cu and P content inbeltline materials and integrated neutron flux. The effect is quantified by means of two variables:USE (upper shelf energy) and RTNDT. Both variables could be obtained from charpy tests. Theaffected area within the vessel is the beltline region.

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The main techniques that allow surveying and monitoring the vessel condition are: Plantinstrumentation and transient register, surface inspection, UT inspection, eddy current testing(mainly for thimbles tubes), vessel surveillance programme, integrity analysis and acousticemission (loose parts). In addition, other techniques which could be used in the future are: Integritymonitors, X-Ray diffraction, fatigue fuses and potential drop.

Surveillance programs based-on capsules are used to evaluate the irradiation embrittlement. SpanishNPPs are following designer country codes and norms to apply the above mentioned surveillanceprograms. In particular, N.R.C. issued a Generic Letter 92-01 “Reactor Vessel Structural Integrity”,10 CFR 50.54(f). This G.L. is aimed to:

� Verify the surveillance program of reactor vessel material: 10 CFR 50, App. H and ASTM E-185.

� Evaluate and limit damage, especially within beltline and flange: 10 CFR 50 App.G.

� Control irradiation effects and applying the results to predict embrittlement level and determinethe credibility of data coming from capsules: G.L. 88-11.

� Assure the fulfillment of toughness requisites against pressure thermal shock (PTS): 10 CFR50.61.

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Theoretical evolution of RTNDT and decreasing of USE are required in NRC G.L 92-01 according tothe methodology of Regulatory Guide 1.99 Rev. 2. Appendix G of 10 CFR 50 and 10 CFR 50.61provide the maximum and minimum values acceptable for both parameters.

In case of actual data availability, provided by the analysis of extracted capsules, the real values ofthe parameter should be calculated. Then, with a minimum of two values, it is possible toextrapolate the behavior of RTNDT and compare with the limits provide by the norms.

245

The requirements for the pressure and temperatures limits, P-T curves, are given in Appendix G of10 CRF 50, and the corresponding methodology in Appendix G of Section XI of the ASME Code.

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� Core re-design: Reduction of the neutron flux by using core-loading pattern of low leakage type.Applied in several Spanish reactors.

� Vessel anneal, if needed.

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In addition to irradiation embrittlement surveillance programs above mentioned, there are additionalactions within the Spanish NPPs / electricity sector focussed on the optimisation of irradiationembrittlement surveillance:

� UNESA Database on PWR surveillance data.

� Contribution to the IAEA Coordinated Research Programs III and IV.

� Reconstitution of specimens and Master Curve evaluation (under consideration by UNESA-CSN).

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Irradiation embrittlement is one of the main problems presented in the life extension process of avessel. That is the reason NRC generates an “Information Notice”, IN- 90-52 recommending theplants to preserve broken Charpy specimens already used. Such an action allows to reconstruct anddata use to keep information uncertainty to a minimum level in what refers material embrittlementof a specific material. Then it is possible to reduce the conservative margins adopted in irradiationcalculation.

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The degradation mechanisms typically affecting steam generator are the following:

� Denting: plastic strength of tubes. Denting affects penetration diameters of tube sheets that arereduced and closed by corrosion products.

� Intergranular corrosion: it affects non-stabilised austenitic steal and Inconel 600 MA. Theintergranular corrosion may present different kinds such as intergranular attack, SCC, andintergranular penetration.

� Pitting: it affects the secondary side of steam generator tubes in the sludge contacting areaaccumulated on tube sheets. This mechanics may produce drilling on materials without anymaterial properties losses.

� Wastage or thinning: external surface tube corrosion. It affects areas with high level of sludgeon tube sheets.

� Fretting: Mechanic effects produces this kind of corrosion; there are not electrochemical factorsinvolved.

� Fatigue: Associated with cyclic service.

� Erosion-Corrosion: affecting inside (corrosion) and outside (erosion) of the tubes.

� Fouling mechanics: it assists corrosion mechanisms.

246

Most common degradation mechanisms affecting Spanish steam generator are PWSCC (PrimaryWater Stress Corrosion Cracking) in the roll transition zone and ODSCC in the tubes support plateintersection, as well as fretting due to AVB (anti-vibrations bars).


The methods to determine degradation mechanics effects are based on inspection techniques.Following are stated the main methods:

� Eddy Current: It is the most extended technique and allows determining numerous parametersrelated to structural integrity of the tubes such as thickness or defect typology. It is applied todetect denting, intergranular corrosion, pitting, wastage, fretting, fouling and erosion-corrosion.

� Ultrasonic testing and radiography: Both methods allow detecting internal cracks and sizingthem. Applied in nozzles and shell welds. Degradation mechanisms studied by UT are mainlyintergranular corrosion, fretting and fatigue.

� Dye penetrant and magnetic particle testing: Applied to detect and size surface cracks producedby any degradation mechanism mainly within the SG shell and welds.

� Visual inspection, to determine the status of internal components.

� Hydrostatic test: Applied to detect full thickness cracks. It is limited due to fatigue risksinvolved within this test.

� Helium leakage: This method is used to demonstrate the hermetic status of a component,including secondary and primary sides of SG.

� Parameters (pressure drop, temperature, hydrogen, etc.) monitoring: Applied to knowgeneralised corrosion risk and fatigue usage by cycle counting.

� Sludge analysis, to determine the corrosion risk and susceptibility.

� Metallographic analysis of tubes to determine the dominant degradation mechanisms.


The acceptance criteria normally used in Spanish NPPs are the following:

� Tubes are plugged when thickness loss rise 40%.

� Maximum acceptable plugged tubes is 10%.

Lifetime prediction methods more common used are based on the evolution of affected and pluggedtubes for each degradation mechanism. Different predictive models have been developed, usingmathematical and statistical approaches and considering historical results from Spanish steamgenerators, in order to predict the evolution of tubes plugged and numbers of tubes to be plugged innext outage.

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Mitigation techniques that can be applied to prevent ageing effects are:

� Secondary chemistry control: The application of this technique leads to minimising environmentaggressivity, reducing corrosion risk. It could be complemented by the surveillance ofparameters as conductivity, pH, cupper content, etc.

� Transient minimisation. Clearly, it is aimed to reduce fatigue risks. The appropriate surveillanceof transient facilitates lifetime calculations.

� Operating temperature decrease: This measure contributes to minimise transient severity as wellas decrease the effect of degradation mechanism assisted by temperature. It must be evaluatedthe associated performance losses.

247

� Sludge areas cleaning: It contributes to obtain better performance and avoid corrosion risks.

� Cupper components replacement: This action is aimed to minimise chloric compounds, air andcupper entrance.

� Tube sleeving: protect tubes affected areas.

� Tube plugging: It prevents all degradation mechanisms, especially PWSCC. It should beevaluated performance reductions.

� Plug replacement: in case of Inconel 600 plugs, due to the problems associated to this material.

� Stress relaxation: To reduce SCC risk.

� SG replacement: last measure.

Most common mitigation methods used in Spain, before the replacement of steam generator in fourunits have been: chemical polishing, sludge lancing, use of coordinated Li/B primary waterchemistry and thermal treatment of tube-U for stress relaxation.


Steam generator have been replaced in four units: Almaraz I and II and Ascó I and II. Before thereplacement, additional studies and actions were performed:

� Maximum acceptable plugged tubes criteria was changed, from 10% to approximately 18%,based-on manufacturer studies.

� Specific plugging criteria were determined for each type of degradation, in order to guaranteethe steam generator structural integrity and the prevention of in-service leakage.

In addition, other minor replacement have been produced, as example the AVB (anti-vibrationsbars) at Vandellós II NPP to avoid fretting degradation.

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Operating experience shows that some plants (Farley2, Tihange and Genkai) have had totalpenetration cracks in non-isolable sections of piping that are connected to the Primary System. Theanalysis of these failures show that these faults are basically caused by thermal stratification inpiping caused by leakage in isolation valves.


The following actions have been established in order to assess the potential impact of thermalstratification:

� Identification of piping sections that may be subjected to thermal stratification and pipe swingscaused by valve leakage.

� Inspection of welds, heat affected zones (HAZ) and stress concentration zones.

� Establishment of a mitigation plan to prevent faults in the identified sections.

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� Instruments with sufficient gain that are capable of distinguishing cracks of other insignificantreflector elements.

� Transducers with rays in different angles.

248

� Records of all indications, including minimum amplitude signals.


1.1.1.2.1.1. Prediction

Due to the small number of incidents due to cracking caused by thermal stratification, it isnecessary to develop a prioritisation process to allow selection of the most critical locations thatwill consequently be subjected to UT inspection. The criteria used are:

� Probability of leakage through the non-isolable section during normal operation.

� Value of temperature difference on both sides of the valve.

� International operating experience regarding leakage or cracking.

� Sensitivity to thermal stratification (function of the piping inside diameter).

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Cracks detected during the inspection are accepted in accordance with the ASME XI criteria, whichimply an assessment of the increase of fatigue in the crack detected, comparing it with thecalculated size of the critical crack in the corresponding section.

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The results of the inspections have shown that the elements inspected are in good condition, and theadditional stress and fatigue analyses that consider thermal stratification transients give anacceptable result of the current piping layout and support. Consequently, there has been no need fora medium long-term solution.

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Erosion-corrosion is a well-understood cause of degradation in pipes. It causes loss of thickness inpipe walls, resulting in degradation of pipe components in carbon and low alloy steel lines untilthey no longer operate safely. For this reason there is widespread concern that erosion-corrosionshould be monitored.

It has become clear that erosion-corrosion occurs when a flow of water or wet steam erodes ordestroys the oxide layer that protects the surface of the pipe; new oxide form and exposed materialdissolves. The following parameters govern the erosion-corrosion process:

� Thermohydraulic conditions in the fluid.

� Water chemistry.

� Chemistry of the pipe material.

� Geometry of the component and exposure time.

1.1.1.3. Surveillance / Periodic Testing

Elements vulnerable to Erosion/Corrosion are identified through the evaluation of the above-mentioned parameters that characterise this degradation. After identifying these elements,inspections are defined that check their condition by means of thickness measurements usingultrasonic techniques (UT). The determination of the frequency and scope of these inspections isbased on the results of previous inspections and the evaluation of the characteristic E/C parameters.

249

1.1.1.4. Ageing Prediction Criteria

1.1.1.4.1.1. Prediction

Several researchers have produced mathematical models of the influence of each degradationparameter on material loss. These models have allowed the formulation of computer codes that arenow widely used.

An initial selection of lines and components most subject to degradation by erosion-corrosion isbased on a rough analysis. There are two ways to make this selection:

� Use a computer code to predict the erosion rate.

� Draw on experience, engineering data and expert judgement. This is the practice in many plants.


The acceptance criteria are taken from the ASME code. The next inspection and the date to replacethe damaged element can be determined through the analysis of inspection results and the use of apredictive programme. When the inspection results show that the minimum thickness will bereached before the next inspection, and there are replacement limitations, the element is reevaluatedapplying ASME Code, i.e. Code Case N597.

1.1.1.5. Mitigation of ageing Effects

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1.1.1.7. Maintenance Programs/ Component Repair/ Replacement/ Improvement

Replacement of the affected elements. Replacement normally involves a change of material, beingthe new elements of high alloy steel or stainless steel.

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Spanish NPPs have been defining the ageing control programmes for concrete structures since the1990s. Said programmes are aimed at monitoring the condition of these structures in order toanticipate ageing by taking measures to monitor it, if necessary, throughout the lifetime of theplants. These structure condition monitoring programmes have been completely generalised withthe implementation of 10 CFR 50.65 “Maintenance Rule”

1.1.1.8. Surveillance / Periodic Testing

The structure monitoring programmes follow the general outline below:

� Identification of critical structures.

� Determination of the type of inspection for each critical structure (initially differentialsettlement measurement and visual inspection, and samples, mechanical test and chemicalanalysis if needed). Determination of frequency of inspection (3, 5, or 10 years).

� Determination of repair methods according to the type of damage and degradation mechanism.

1.1.1.9. Ageing Prediction Criteria


The results of the visual inspection are evaluated by civil engineering experts. The results of themechanical test and chemical analysis are evaluated with respect to the properties included in thetechnical specification for the concrete.

250

1.1.1.10. Maintenance Programs/ Component Repair/ Replacement/ Improvement

The concrete and/or steel frame repair method is defined according to the type of damage and thedegradation mechanism.

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The wiring of equipment are under several degradations phenomena:

� Thermal Ageing: it is caused by environment temperature and wiring heating due to Ohmiclosses.

� Humidity: higher humidity reduces wiring dielectric properties and insulation.

� Radiation: it causes degradation in mechanic and dielectric insulation characteristics.

� Electric gradient: it produces dielectric degradation.

Electrical cables installed in the Spanish NPPs belongs to different manufacturers, materials andage. Thus the use of generic methodologies for age assessment is not feasible.

Plants more concerned by cable ageing are those which belong to the first generation of reactorswhich have been operating in the 25 years range. In these plants, due to the old technology of theoriginal cables designs and the accumulated operation of the plant cable ageing is an issue which isbeing addressed at this moment.


The most common methods to perform wiring surveillance are electric and mechanical testing andvisual inspection. Some of them are:

� Insulation resistance measurement / Polarisation index.

� Tg delta measurement and capacitance test (MT).

� Partial discharges testing (MT).

� Elongation and tensile strength.

� Indentation test.

� Visual inspection to detect damage in the wire insulation, ending and terminals.

Limited systematic activities regarding cable testing and surveillance have been addressed by theSpanish plants up to the recent years.

Safety related cable ageing was addressed by the environmental qualification programsimplemented as required by IEEE standards. These cables have been subjected to acceleratedageing test, addressing thermal and radiation induced degradation and then subjected to LOCAsimulation test. The results from these programs provided an in-sight of cable capability during thelife of the plant.

In recent years a systematic approach to the management of cable ageing has been started bydifferent plants. Some of these plants are implementing electrical inspection activities in importantcable ( in terms of safety or availability) using the ECAD system and methodology. One plant hasimplemented cable sample in hot spots (temperature and irradiation) for periodic inspection of cablematerials degradation.

Our oldest power plants are starting to define systematic ageing managing programs. In additionelectrical cables are being included as a significant item in the life management program beingdeveloped at the moment by the pool of the nuclear power plants.

251


Ageing prediction methodology will be started to be implemented within the life managementprogram being developed. This methodology will be based on the criteria developed abroad ,basically in the US, where an ageing model which relates cable degradation to different testparameter (tensile strength, indentation, Oxidation Induction Temperature, etc) is established.

1.1.1.11. Mitigation Of Ageing Effects

Keep to a minimum cover material and insulation temperature in values, is the way to mitigate theageing effects. The actions to avoid the premature thermal ageing are:

� Intensity reduction in normal operation.

� Apart wiring near to heat sources.

� Modify the wire situation in order to improve the refrigeration.

� Selective monitoring in order to survey temperature and voltage.

� In case wire must to be changed due to environment degradation it is necessary to reduce theenvironment conditions.

� Check environment qualification test in order to detect wires that must to be changed.

Nevertheless, significant cable ageing is normally addressed by cable replacement. In some cases,typically hot-spots, cable re-routing or protection is implemented to prevent further degradation.


In the older plants extensive efforts has been performed during years to replace old cables whichwere considered as age-sensitive.

Nowadays approach to cable ageing management focuses in an early identification of cabledegradation in order to be able to implement corrective measurements, normally cable replacement.

252

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111/ Regulatory policy and requirements Spanish Nuclear Safety Authority (CSN) Regulatory Guide 1.10, Operating Licensee, Periodic Safety Review (PSR)

112/ AMP policy Spanish NNP approach, which is described in UNESA documents “NPP Life Management in Spain.- Description of Methodology ”and “Licensing and Regulatory Aspects on Lifetime Management in Spain” included in Sections 4.3 and 5.3 respectively.

113/ International guidance IAEA Safety Series No.15, 10CFR50.54

114/ Scope of AMP The UNESA methodology includes a procedure for the SSC selection: “Guideline for Selecting Critical Components”, and for eachcomponent or group of components a detailed dossier allow to select the pairs area/component and degradation phenomena whichare most critical.

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121/ Policy of the N.S.A. on AMPs Refer to result of "Nuclear Safety Group" analysisThe RG 1.10 requires a surveillance of ageing issues. To obtain the renewal of the Operating Licensee, Lifetime Management Plansare mandatory.

122/ Additional regulatory requirements andguidance by the N.S.A.

CSN can ask for some plant specific requirements on AMP after Periodic Safety Review. When the plant life be next to the designlife (40 year), additional requirements could be required. Currently NSA does not define additional requirements.

123/ The AMP policy document of theowner/operator

Following UNESA methodology, a specific Lifetime Management Programme is generated for each NPP.

124/ Available international guidance and goodpractice

IAEA, US-NRC, EPRI, INPO, WOG (PWR), BWROG.

125/ The scope of the AMP Selected SSC, considering the high susceptibility to degradation phenomena. The main activities include the degradationphenomena studies and the maintenance practices evaluation.

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211/ AMP organization and programdescription

Specific documents for each plant, including Lifetime Management Programme description, responsible and organization. Someplants have a specific procedure on AMP organization.

212/ Resources:

(a) human

(b) financial

(c) tools and equipment

(d) external

NPP Life Management Programme includes multidisciplinary teams with engineering companies support.

Utilities/NPP funds for its specific Programme. Electricity Sector/UNESA coordinates common NPP methodology and approaches,and addresses R+D programmes for projects on Materials, supporting the NPP Life Management Programmes.

Databases, inspection/surveillance equipments, maintenance tools, etc. A specific support tool: “Integrated Lifetime ManagementSystem (SIGEVI)” is being developed by UNESA.

External resources are from engineering and inspections companies, main equipment suppliers, etc.

213/ Provisions for understanding SSC ageing Base documents generated by UNESA. Additionally, specific documents developed by supplier working groups are being used, i.e.WOG and BWROG.


221/ Organizations participating in AMP andtheir interfaces.

UNESA, CSN, Utilities, NPP staff and external consultants (engineering, suppliers, services and inspection companies).

222/ The division of responsibilities CSN (safety), UNESA (methodology and coordination), NPP owners (AMP responsible) –see figures 2-1 and 2-2–, Engineering,Inspection, Supplier companies (external support).

223/ Criteria used to determine the AMorganization

Specific for each NPP (Plant staff availability, experience in different areas covering, among others, equipments maintenance andoperation, degradation phenomena, etc.).

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224/ Resources allocated to the AMorganization

Specific for each plant (Experts for different services covering main degradation phenomena and components).

225/ The independance of AMP funding Methodology (Electricity Sector/UNESA), Plant specific programmes funded by the Utilities / NPP.

226/ The qualification required for staff Multidisciplinary teams (experts in materials, ageing, analysis, inspection, maintenance, operation, quality assurance, etc.).

227/ The previous experience of AMorganization staff

The expert must have previous experience in the area of their knowledge. Not specific requirements are obligated.

228/ Training program to ensure thecompetency of staff

The training programmes are related to the specific knowledge matter of the expert.

229/ The adequacy of the equipment and tools There are a high number of available tools. Most of them are being optimised for Lifetime Management objectives and integratedwith other related tools.

2210/ Feedback of relevant operatingexperience and research results

The Lifetime Management Programme of each NPP is feed backing the results of the Programme application. Component analysismodifications are examples of the continuous improvement and feedback of results.

2211/ and their application ...

2212/ External expert advice whennecessary

Usually, engineering companies and equipment suppliers are considered when necessary.

2213/ Advice from national andinternational organizations

Good practices, advises, reports and documents from UNESA, IAEA, OECD/NEA, US-NRC, EPRI, NEI, VGB, etc. are consideredin NPP AMP.

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Responsible forActivity … N

Life ManagementProgramme Manager

(Utility-NPP)

Responsible forActivity 1

Responsible forActivity 2

NPPSections

ExternalEngineering Support

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(CSN)

LIFE MANAGEMENTCOMMITTEE(Utility-NPP)

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Life ManagementProgramme Manager

(Utility-NPP)

Technical Managerfor LM Basic Activities

(Engineering Support)



SpecificLM

Activities

Nuclear SafetyAuthority

(CSN)

NPPSections

258

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311/ SSC screening method Procedure included in the UNESA methodology and in the specific LM programme developed by each NPP based on this methodology (UNESA’sdocuments were provided to the Project: “Description of Methodology” and “Guidelines for Selecting and Prioritising Important Components”,summaries of which are in Chapter 5, Section 5.3).

312/ List of SSCs Important SSC selection following the UNESA procedure “Guideline for Selecting and Prioritising Important Components” (example lists areprovided in Appendix 2, Section 4: “Lists of Important Components.- Results from UNESA Methodology”).

313/ Operational procedures Each selected component or group of components is analysed in a detailed dossier, covering: component design and operation characteristics,maintenance and inspection history, degradation phenomena affecting, evaluation methods, and possible measurements for degradation control andmitigation.

314/ Surveillance Each selected component or group of components is analysed in a detailed dossier, covering: component design and operation characteristics,maintenance and inspection history, degradation phenomena affecting, evaluation methods, and measurements for degradation control andmitigation. In order to increase SCC surveillance, specific equipment and programmes have been designed, i.e., vessel head penetration inspectiontool, vessel bottom penetrations inspection device, predictive maintenance of steam generators developments, etc..

315/ Assessment Each selected component or group of components is analysed in a detailed dossier, covering: component design and operation characteristics,maintenance and inspection history, degradation phenomena affecting, evaluation methods, and measurements for degradation control andmitigation.

316/ Maintenance As key stage of the UNESA methodology, an important activity is the Maintenance Evaluation to assess the degree to which the degradationmechanisms are covered by the maintenance, inspection and testing practices currently implemented at the NPPs. The procedure developed for theMaintenance Evaluation allows all the activities related to maintenance (preventive, corrective, testing, inspections, operating shift plant walk-throughs, etc.) to be analysed in an integrated manner as a means of controlling and mitigating ageing.

317/ Equipment qualification program procedures The Spanish Nuclear Safety Authority (CSN) requires NPPs to implement equipment qualification programmes. These follow the regulations ofthe plant design origin country (i.e., US design plants use IEEE-323 for environmental qualification and IEEE-344 for seismic qualification).Equipment qualification programmes are considered in the NPP Periodic Safety Reviews and Lifetime Management Programmes.

318/ Data collection and record keeping Each plant use its own procedure and system to collect and record data, using plant process computer, maintenance and inspection databases,chemistry analysis database, etc.

319/ Spare parts Policy of spare part management is plant/utility specific. At this moment, groups of plants with unified management are establishing an unifiedspare part management policy, in order to optimize resources and minimise stocks based on spare parts shares between plants. Spare partmanagement is considered in the NPP Periodic Safety Reviews and Lifetime Management Programmes.

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321/ SSC screening methodology document,the list of SCCs and examples

The plant SCCs are selected based on different criteria, including Safety, Availability, Replacement and cost, Service conditions,Regulatory considerations, Operating history, Effectiveness of plant programmes, among others (summary on UNESA SSCscreening methodology was provided in Project Task 3: “Procedure for Selecting and Prioritising Important Components).

322/ Safety margins and/or acceptance (fitnessfor service) criteria specified for the SSCs

Refer to results of "Technology Group Analysis". In accordance with regulations of the plant design origin country. Specific studiescould be required in certain cases. Typical criteria examples are:

• Admissible design transient

• Minimum wall thickness

323/ Ageing assessment methodology fo SSCsand examples

Refer to results of "Technology Group Analysis". The assessment includes reports/dossiers on degradation phenomena studies foreach component or group of components (vessel, pressurizer, steam generators, turbine-generator set, containment, tanks, piping,pumps, heat exchangers, valves, motors, etc.). Examples are described in UNESA documents provided in Project Task 3:“Description of methodology” and “Licensing and Regulatory Aspects on Lifetime Management in Spain”.

324/ Operating procedures for SSCs Operating procedures are adapted considering design data; i.e., transient reduction, corrosion mitigation by chemistry control, etc.

325/ Plant surveillance program The surveillance programmes are plant specific.

326/ Inspection and surveillance procedures The procedures used are plant specific and consider the best available techniques. Inspection are adapted considering plant historyand state-of-the-art methodology: As examples:

• Ultrasonic test for RPV welds and BWR shroud.

• Eddy current test for SG tubes.

• Specifically designed equipment for RPV head bolts.

Surveillance of operating parameters, pressure, temperature, valve position, etc., is used for automatic fatigue monitoring in severalNPPs. Vibration and noise detection allows to minimize damage on main pumps and other primary components, etc.

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327/ Operational limits and conditions controllingthe rate of ageing

Refer to results of "Technology Group Analysis". Only if excessive degradation has been detected and it is necessary to mitigate it. Operatingprocedures limits for pressure temperature, chemistry parameters, etc. and preventive maintenance, including refurbishment, mitigate SSCdegradations.

328/ Maintenance programs Several plants have reviewed their maintenance programmes in order to improve frequencies in preventive maintenance, reducing preventiveactuations and increasing predictive maintenance.

329/ Preventive maintenance programs

3210/Schedule of minimum preventive maintenance

3211/Basis for adjusting testing, surveillance andmaintenance

Methodologies used for maintenance evaluation are in UNESA methodology for NPP Lifetime Management. Also, RCM. is implemented in mostof the plants.

3212/The changes to the AMP in response tounanticipated ageing phenomena

AMP is in continuous process of review, improvement and feedback.

If problems are plant specific, the plant adopts the adequate measurements to evaluate the incidence and to control and mitigate them.

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3213/Data collection and record keeping system Conventional methods and other specific for lifetime management (fatigue monitoring, SIGEVI, etc.). Spanish NPPs are required to prepare andsend to the Spanish Nuclear Safety Authority (CSN) different license documents and information, which need to be collected and recorded inappropriate way. To manage the plant SSC and the associated documents, databases for configuration control are used by the NPP as requirementsof Quality Assurance.

3214/Maintenance histories, including: NPPs maintain history records, including information on faults reports, repair, replacement and refurbishment performed. Most of them usesoftware programmes for maintenance management.

3215/Procedures and mechanisms in place tomaintain EQ

According to the licensing requirement US-NRC 10CFR50.49 for environmental qualification. NPPs have specific procedures to maintain EQbased on equipment qualification reports. In accordance with the conclusions of periodic safety reviews (RPS), the EQ maintenance programmecould be modified, i.e. if environmental conditions change.

3216/Systematic analyses and corrective actiontaken

Several plants follows the US-NRC 10CFR21, which requires to perform a fault cause analysis and to distribute it to other plants, suppliers, etc.which could experience the same failure mechanisms.

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411/ Physical condition of SSC

412/ EQ established and maintained

413/ Performance indicators

The understanding of the group in this chapter is included in the AMAT guideline to record during the audit the informationobtained.

This is not adapted to the review performed by the 3 Utilities (UNESA / TRACTEBEL and EDF) now.

Corresponding information is already given in activity.


421/ Major findings, conclusions andrecommendations resulting from plantwalkdowns

422/ Special attention given to certain SSCs

423/ Degree to which condition and/orfunctional indicators conform to theacceptance criteria

424/ Records of ambient environmentalconditions

425/ Records of system parameters includingtransients, trends and deviations

426/ Qualification reports or other documents

427/ Statistical information about failures

428/ Routine reports from the maintenance andsurveillance organizations

429/ Trends of AMP process indicators

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511/ Self-assessment program Each NPP has established its own process to review, update and control the AMP. In the UNESA’s Working Group on NPP Lifetime Management& Materials, NPPs exchange information, agree common projects, and follow the respective Life Management Programmes.

512/ Peer reviews No yet between Spanish NPPs on ageing management programmes.513/ Comprehensive reviews Spanish Nuclear Safety Authority (CSN) has audited the UNESA’s Methodology for NPP Life Management, with favourable result. Also, CSN

audits the individual NPP Life Management Programmes.

514/ Continuous improvement process They are foreseen in the NPP Lifetime Management Programmes.��5HYLHZ�7RSLFV�

521/ Regulatory requirements for the AMP review Requirements, including Operating License and PSR, are described in UNESA document provided to the Project: "Licensing and RegulatoryAspects on Lifetime Management in Spain" (summary in Chapter 4, Section 4.3).

522/ Policy for the AMP review and improvement Plant specific. Considered in the plant AMP and not in the methodology.

523/ Performance indicators Plant specific. Considered in the plant AMP and not in the methodology524/ Updates of performance indicators

525/ Records of self-assessments Plant specific. Considered in the plant AMP and not in the methodology.

526/ Plans and procedures of self-assessment Plant specific. Considered in the plant AMP and not in the methodology.

527/ Capability of the NPP owner operator toevaluate AMP effectiveness

It is understood that it usually requires external expert support.

528/ Records of peer reviews529/ Independence of the peer review teams5210/Results of the peer reviews5211/Effectiveness of corrective actions or

improvement programs5212/Records of any comprehensive reviews

5213/Roles of the operator and national regulatoryauthority

Spanish Nuclear Safety Authority (CSN) audits the NPP Lifetime Management Programmes, specifically or as a part of Periodic Safety Review(PSR).

5214/Procedures of the comprehensive AMP review5215/Corrective actions arising

5216/A commitment of all members of the NPPstaff to continuous improvement

Each NPP Lifetime Management Programme involves appropriate staff of all the plant sections, coordinated by the LM Manager. A directcommitment exits from the top NPP/Utility Management, being more strong with the particular NPP age.

5217/Relationship between the periodic safetyreview program and any comprehensive AMPreview

Described in the UNESA document provided to the Project: "Licensing and Regulatory Aspects on Lifetime Management in Spain" (summary inChapter 4, Section 4.3).

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111/ Regulatory policy and requirements The operating license of each NPP

112/ Ageing Management Program policy Refer to project "Continuous Operation of Belgian NPPs".

113/ International guidance International guidance and good practices were considered in order to propose an integrated project (See 112/).

114/ Scope of Ageing Management Program See 112/


121/ Policy of the Nuclear Safety Authority onAMPs


Refer to results of "Nuclear Safety Group" analysis. The license of each NPP, which requires the Utility to reassess the safety of theplant after each 10-year, specifies that the items to be handled during each decennial review are identified in a report written incommon by the Utility, its Engineering Organisation, and the technical support organisation of the safety authorities. Plant Lifemanagement is thus integrated in a permanent review process. However, particular concerns are also addressed on a specific basis(e.g. potential for cracks in the reactor pressure vessel head).


There is no predetermined lifetime for a nuclear power plant either license life or either design life. The Belgian nuclear powerplants will be kept operational as long as they can operate safely and economically.


See 113/

125/ The scope of the AMP The current ageing management program focuses on the passive safety-related components and the non safety-related componentsbut important for the availability of the plant; since specific programmes exist for active safety-related components: maintenanceprograms were established to define actions to perform in order to guarantee the integrity and the availability of these equipmentsduring the exploitation of the NPPs, special programmes for procurement of qualified spare and replacement parts and for definingstrategies for replacement of pieces of equipment which have become obsolete for commercial or technological reasons.

264

CE. DE XI "Ageing" AGEING PROGRAM ASSESSMENT / REVIEW CHART(refer to AMAT Guideline)

AREA OF REVIEW: ��25*$1,=$7,2175$&7(%(/

��'RFXPHQWDWLRQ�211/ AMP organization and program description A specific plant life management project, named "Continuous Operation of Belgian NPPs" has been set up to summarise the knowledge gathered

on the various topics related to Systems, Structures, and Components ageing and to summarise the situation of each nuclear unit with respect tothese ageing phenomena.Those summaries are reported in the Equipment Ageing Summaries.

212/ Resources: (a) human(a) financial(b) tools and equipment(c) external

No specific approach for the project "Continuous Operation of Belgian NPPs". Resources are managed as for other Tractebel projects and thisproject is performed according to the internal QA procedures of Tractebel (as for other projects).

213/ Provisions for understanding SSC ageing See 212/��5HYLHZ�7RSLFV�221/ Organizations participating in AMP and their

interfaces.

222/ The division of responsibilities

The main organisations involved are:- The Federal Nuclear Inspection Agency, responsible for the inspection and surveillance of nuclear activities in Belgium;- The mandated private companies, responsible for making safety analyses, for carrying out all permanent control tasks, and for monitoring the

activities of operators.- Electrabel, the electric utility, responsible for operating the plants.- Tractebel, the engineering organisation, responsible for the design, construction and engineering support to the operation (including periodic

safety evaluation), and is in charge of co-ordination.- SCK•CEN, The Belgian Nuclear Research Center and Laborelec, the Research laboratory of the utilities, responsibles for testing.At each site of Electrabel, the management and the concerned sections (i.e. safety section, maintenance section, ...) realise most of the activitiesrelated to the plant life management based on the informations from the concerned sections of the site and the advises from the Nuclear SafetyAuthorities. The management of Electrabel takes then the decision and defines the basic rules to be followed by the utilities taking into account thestrategy of Electrabel. The engineering organisation (Tractebel) is in charge of the support and the co-ordination of the plant life managementactivities.


See 212/

224/ Resources allocated to the AM organization See 212/225/ The independance of AMP funding226/ The qualification required for staff227/ The previous experience of AM organization

staff228/ Training program to ensure the competency of

staff229/ The adequacy of the equipment and tools2210/Feedback of relevant operating experience and

research resultsSee 212/

2211/and their application...2212/External expert advice when necessary See 212/2213/Advice from national and international

organizations

265



AREA OF REVIEW: ��$&7,9,7,(675$&7(%(/

��'RFXPHQWDWLRQ�311/ SSC screening method312/ List of SSCs

The screening method and a list of passive safety-related components are given in the internal documents of the project "Continuous Operation ofBelgian NPPs". Information regarding to those components (e.g. surveillance, acceptance criteria, remedies, ...) are summarised in the EquipmentAgeing Summaries.

313/ Operational procedures314/ Surveillance315/ Assessment316/ Maintenance317/ Equipment qualification program procedures318/ Data collection and record keeping319/ Spare parts

Due to the high quality standard applied to the design, operation, and maintenance of NPPs, many aspects of plant life management have beenincorporated in the everyday management of the plants since the beginning of their life. These aspects include: the design, quality assurance, inservice inspection, monitoring, testing, preventive maintenance, re-qualification, replacement, and periodic safety reassessments.These aspects are well documented and, for passive safety-related components, the major references are summarised in the Equipment Ageingsummaries (See 311/)

��5HYLHZ�7RSLFV�321/ SSC screening methodology document, the list

of SCCs and examplesSee 311/

322/ Safety margins and/or acceptance (fitness forservice) criteria specified for the SSCs

Refer to results of "Technology Group Analysis". See 313/ Criteria are given in the design documents (e.g. the number of transients experienced bya component cannot exceed the number of occurrences assumed in the qualification analysis) or are determined by specific analysis (e.g. minimalligament length for cover head penetrations).

323/ Ageing assessment methodology for SSCs andexamples

Refer to results of "Technology Group Analysis". See 311/

324/ Operating procedures for SSCs See 313/ e.g. improvements of the water chemistry to control the Steam Generator degradation325/ Plant surveillance program326/ Inspection and surveillance procedures

See 313/ e.g. steam generator tube inspection by Eddy Current and/or Ultrasonic examination by sampling at each refuelling outage.


Refer to results of "Technology Group Analysis". See 313/ e.g. pressure and temperature limits for heat-up and cool down (RPV)

328/ Maintenance programs329/ Preventive maintenance programs3210/Schedule of minimum preventive maintenance

See 313/ e.g. control of degradation due to erosion-corrosion by wall thickness measurements of the most susceptible elements which theinspection frequency depends on the severity of the degradation in order to replace in time the thinned pipe portions


All Classified pressure retaining components are inspected according to the requirements of the ASME Code Section XI. Moreover,complementary or voluntary inspections are decided by the operator on classified and non-classified components according to variable aspectsaffecting the availability and the conventional security of the plant or depending on world feedback or experience.


See 313/ e.g. local monitoring was implemented to characterise thermal stratification transients

3213/Data collection and record keeping system3214/Maintenance histories

Results of the maintenance programs are kept and are used, among other things, in order to optimise the preventive maintenance programs (using amethod based on the Reliability Centred Maintenance).



e.g. for electrical and I&C components, during quality audits which occur every 3 years, manufacturers are questioned on the evolution of their 1Equalified products and on the obsolescence of their products. This audit allows to check if the material presently offered still meet the requirementsof the original qualification.

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511/ Self-assessment program

512/ Peer reviews

513/ Comprehensive reviews

514/ Continuous improvement process

See 212/


521/ Regulatory requirements for the AMP review No specific regulatory requirement for the AMP review

522/ Policy for the AMP review and improvement See 212/


524/ Updates of performance indicators

See 212/

525/ Records of self-assessments

526/ Plans and procedures of self-assessment

527/ Capability of the NPP owner operator toevaluate AMP effectiveness

See 212/

528/ Records of peer reviews

529/ Independence of the peer review teams

5210/Results of the peer reviews

5211/Effectiveness of corrective actions orimprovement programs

See 212/

5212/Records of any comprehensive reviews

5213/Roles of the operator and national regulatoryauthority

5214/Procedures of the comprehensive AMP review

5215/Corrective actions arising

5216/A commitment of all members of the NPPstaff to continuous improvement

See 212/

5217/Relationship between the periodic safetyreview program and any comprehensive AMPreview

Plant-Life management is integrated in the periodic safety review. However, specific concerns are addressed on a specific basis (e.g. potentialcracks in reactor vessel head)

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111/ Regulatory policy and requirements See Safety Group

112/ AMP policy Refer to « Service life and ageing of pressurized water power plants » (EDF ref. 1)Situation end of june 1998

This overall generic program makes use of all skills available within EDF and the manufacturers, and guarantees consistent and complete followup analysis by making use of numerous expert areas. Beyond most pieces of equipment which can be maintained during maintenance activities,there are what EDF consider as non-replaceable components. The former can usually be replaced within a given time period at a controlledradiation exposure and a reduced cost. For instance, at the end of 1997, EDF replaced the steam generators in seven units and twenty four reactorvessel heads.

113/ International guidance IAEA SRS 15 (it has included EDF contributions).

114/ Scope of AMP Refer to (EDF ref. 1).


121/ Policy of the NSA on AMPs


Refer to results of "Nuclear Safety Group" analysis. This policy is defined by official letters and recommandations written after « Groupepermanent » meeting devoted to Ageing. These GP are leaded by French Safety Authority and include expert from all parties involved.


The design of the nuclear steam supply system was based on a technical service life of 40 years, the period taken into account in the safety reports.However, from the point of view of the regulations, French legislation does not specify a time limit for the operation of installations within theframework of the enactment authorising their creation. With respect to the Safety Authorities and the general public, EDF must therefore strive tobe able to justify and to achieve this service life and, if possible, to prolong it in order to benefit to the full from the investments already made.

Three main factors have an influence on the service life of a nuclear power unit:

- normal wear on its components and systems, sometimes referred to as ageing, which depends particularly on their age, their operatingconditions and the maintenance operations performed on them;

- the safety level, which must conform at all times to the safety reference system applicable to the unit and which is likely to evolve according tonew regulations;

- cost-effectiveness, which must remain satisfactory in comparison to other means of power generation.

In this context, obtaining a service life of 40 years depends on the control, on one hand, of a safety level that must conform at all times to thereference system and, on the other, of all the technical and industrial aspects that make it possible to operate the units safely and cost-effectively.

From a technical point of view, this entails gaining an understanding of the problems of ageing, defining and implementing suitable measures tomaintain the performance of the units at their current level. From an industrial point of view, this entails mastering the evolution of the industrialfabric so as to have the necessary skills, know-how and tools on hand when required. In all cases, it is in the interests of EDF to ensure that theindustrial skills exist in France or elsewhere in the world to contribute to the operation of the units, even in the absence of prospects for renewal.

On account of these elements, the overall strategy is founded firstly on the quality of daily plant operation, and subsequently on the followingpoints:

- the ten-yearly reassessment of the safety level;

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- the implementation of two structured programmes so as to ensure that all the technical and industrial actions allowing a service life of at least40 years to be achieved have indeed been put into effect:. the “anticipation and exceptional maintenance” programme;. the “service life” programme.

Obviously, this process and these programmes are the subject of continuous exchange with the Safety Authorities.


AIEA documents: SRS n° 15 : Implementation and review of a nuclear power plan ageing management programme.SS n° 4 AMAT Guidelines : Reference document for the IAEA Ageing Management Assessment Teams.

These documents are not referenced in EDF document (due to their issuing date but EDF specialists have been involved).

125/ The scope of the AMP The aim of the service life programme, initially set up by EDF in 1987, is to understand and anticipate the problems of ageing. This programmeserves as an active “observatory” to ensure that everything is done to achieve the expected service life. Within its scope, all the aspects that have animpact on the service life of installations are reviewed, whether these are purely technical, i.e. associated with the equipment, or of an industrial,economic or regulatory nature.

The service life programme also serves to identify the progress required to gain a better understanding of ageing phenomena and to undertakeresearch and development allowing the link between operating conditions, maintenance and service life of the components to be more clearlyestablished. The programme makes the distinction between:

- two irreplaceable components: the reactor vessel and the containment buildings. A summary document demonstrating the serviceability of 900megawatt PWR vessels, for at least 40 years, is currently being investigated by the Safety Authorities. This document makes provision forcloser monitoring of irradiation behaviour and for consideration of the specific characteristics of each vessel. As regards the containmentbuildings, a service life of at least 40 years has been established overall for the 900 megawatt standardised plant series. The monitoring andactions on the containment buildings of the 1300 megawatt series have to be individualised, given that loss of leaktightness has occurred on theinternal walls of certain containment buildings (Belleville, Flamanville, Cattenom);

- components that can be wholly or partially replaced, sometimes requiring onerous operations that are nevertheless fully mastered nowadays,such as steam generators (already replaced on 7 units) or vessel heads (already replaced on 30 units). The corresponding actions are integratedin the exceptional maintenance strategies while observing the objective of “at least 40 years”.

Obviously, ambitious research and development programmes are dedicated to the investigation of the mechanics and kinetics of deterioration:erosion, corrosion, fatigue, wear, ageing under the effects of heat and irradiation… Indeed, a thorough understanding of these phenomena isrequired in order to optimise strategies throughout all the French plants in service. To confirm this work, a series of expert appraisals has beencarried out on the Chooz A plant (300 megawatt), the first pressurised water reactor to be built in France and decommissioned in 1991 after 24years of service. Furthermore, EDF monitors feedback from foreign nuclear power plants older than the French units attentively and organisescollaboration with the plant operators concerned. In particular, the plants in service in the United States, totalling approximately one hundred PWRunits with an average age 10 years older than those run by EDF, are a major source of feedback on the service life of equipment. The performanceof the majority of these plants continues to improve, which lends weight to the argument that considerable margins exist as regards the end of theservice life of such units. The service life programme also addresses the question of the perpetuity of the nuclear industry, in order to monitor thesituation of industrial organisations that may be both “sensitive” owing to the indispensable nature of their skills for EDF and “fragile” owing tothe absence of new construction for a certain number of years.

270

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211/ AMP organization and programdescription

A committee « Projet Durée de Vie » was set up in 1985 to make synthesis about ageing and life time management.For implementation of modification, the normal process as per any modification is used : based on budget estimation decision toperform is made and activity is done by a Technical Division in charge of the field.

212/ Resources :

(d) human

(e) financial

(f) tools and equipment

(g) external

They are mobilized in the Engineering, Research and Operation Division as per any other matter. No specific approach has beennecessary : decision were taken by the Steering Committee of the Operation Division as per any study or modification decided onEDF units.

213/ Provisions for understanding SSC ageing Research Division is performing a huge program linked to ageing mechanisms.


221/ Organizations participating in AMP andtheir interfaces.

222/ The division of responsibilities

The « PDV » is in a position of advisor for the Production Division Head. It has no specific means except for the expert to attend themeetings. The decision to launch an action (either studies or modification) is decided by the Steerring Committee of Operatingdivision. As such each action is individually managed by the Division in charge.

Organisations participating : see Table 1

Decision making / interfaces : see Table 2


The division of responsibility is the same for AMP as per any study or modification. Some Engineering department have a keycontribution.

224/ Resources allocated to the AMorganization

225/ The independance of AMP funding

226/ The qualification required for staff

227/ The previous experience of AMorganization staff

228/ Training program to ensure thecompetency of staff

229/ The adequacy of the equipment and tools

Resources are managed at Department and Division level depending of the part of the job they are responsible for. The ageingactivity is only a part of their activity and, as such, participates in load and budget evaluation.

Qualification of personnel is also decided at department division level depending on the field of activity and the requested skill asper any other activity.

People in charge of ageing studies are managed, trained in the same way as per any other engineering activity (according to the fieldof activity).

271

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2210/ Feedback of relevant operatingexperience and research results

2211/ and their application...

The feedback is organized by two levels:

- a NPP level: for each incident or failure local Engineering team analyses and corrects if it is a local problem,

- if the problem may be generic, enquiry is extended to similar units and the problem is managed by a corporate level for strategicdecision.

It is the same for ageing related problem for which PDV has an overall coordination mission.

2212/ External expert advice whennecessary

2213/ Advice from national andinternational organizations

Technical divisions are used to ask for external support:

- from original supplier in particular Nuclear boiler supplier,

- other companies in the world in particular the original licenser.

- In addition, EDF participate in many common research programs developped by international organizations such as WANO,FROG and INPO

272

CE. DE XI "Ageing"


AGEING PROGRAM : FRENCH ORGANIZATIONS PARTICIPATING



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311/ SSC screening method The method is based on identification of sensitive equipment depending one ageing possible damage estimate, impact on operationof the plant in term of safety, availability and replacement costs.

312/ List of SSCs A list of sensitive equipment having a major impact on life duration has been devided in changeable or not equipment. The list is???.

313/ Operational procedures They include from the beginning the impact of transient taken into account in design for fatigue analysis.

They are complemented by chemical procedures.

314/ Surveillance EDF has developped a set of Preventive maintenance Programmes based on an optimization process, called "OMF-Structures".

The principles of the OMF-Structures process include "Risk-Based Inspection" concepts within an RCM process. Two main phasesare identified :

- The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of twomajor steps: potential consequences evaluation and reliability performance evaluation. For consequence evaluation, bothquantitative PSA and deterministic aspects are used. For performance evaluation, degradation models are used to select "sensitiveelements" where degradation mechanisms may occur.

- The second phase consists of the definition of preventive maintenance programs for elements that are associated with risk-significant failure modes. The tasks and frequencies are proposed depending on the nature of the element and the degradationmechanism attributes (attributes, kinetics, ...). For high risk failure modes, a probabilistic optimization is to be proposed.

315/ Assessment

316/ Maintenance

317/ Equipment qualification programprocedures

A large program for qualification has been developped:

- it starts from the list of functions to be qualified: this list has been based on functional analysis of incident / accident situationand how to bring the boiler to a safe state.

- Equipment necessary to perform such a function are indentified and each component participating also noted in a list ofequipment to be qualified.

- A list of qualified equipment.

318/ Data collection and record keeping

319/ Spare parts They are managed by a specific division which distribute information to sites. Spare parts linked to qualified equipment areidentified.

276

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321/ SSC screening methodology document,the list of SCCs and examples

- The list of sensitive SS components (from life duration point of view) is attached to EDF ref. 1 and regularly updated,depending on result of analysis (ageing speed – Faisability of changing the equipment).

322/ Safety margins and/or acceptance (fitnessfor service) criteria specified for the SSCs

Refer to results of "Technology Group" analysis :

- Criteria were given in the design documentsi.e. : number of transient for 40 years operationminimal thicknesshypothesis on toughness.


- Criteria have been determined later on by specific analysisi.e.: analysis of cracks propagation in material (inconel)

analysis of embrittlement of austenitic cast iron products.

323/ Ageing assessment methodology for SSCsand examples

This chapter shall be linked with technological group analysis, refer to ref. 1: the attached sheet description for the main sensitiveequipment the trends and the methodology to assess them:

- i.e. : Reactor vessel

324/ Operating procedures for SSCs Initial set of procedures takes into account design criteria such as profile of transient, operation domain; chemistry of fluids.

i.e.: - thermal shock on reactor vessel,

- material (i.e.: low C content stainless steel) adapted to primary corrosion.

After commissionning some additional studies have given complementary criteria and operation has been adapted:

i.e.: - temperature of reactor vessel upper part lowered to reduce cracking of inconel weldings,

- smooth start-up of diesel generator set to reduce wear of internal parts.

325/ Plant surveillance program PBMP (Programme de Base de Maintenance Preventive). This programs, issued by equipment, give the maintenance tasks, theirfrequency. They are the result of the methodology given in § 314.

277

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326/ Inspection and surveillance procedures - Inspection is developped and adapted to the forecast type of wear:

I.e.: - ultrasonic test of reactor vessel weld to detect crackings,- magnetic test for steam generator tube cracking,- ultrasonic test for thickness reduction of carbon piping under erosion corrosion.

- Surveillance (during operation): in some particular cases equipment to check some parameters have been installed:

I.e.: - vibration surveillance of reactor vessel internals to detect any damage,- electrical parameters of motorized valve motors to detect any increase in operating efforts,- computerized transient identification system (SYSFAC).



Initial set of operational limits and conditions were settled in accordance with design: refer to operating procedures chapter.

After commissionning some additional limit have been decided to take into account additional studies:

i.e.: - ageing of non metallic parts has been studied and systematic changes are included in maintenance procedures for membranes ofpneumatic valves.

328/ Maintenance programs

329/ Preventive maintenance programs

3210/Schedule of minimum preventive maintenance


See 325

See 325

See 325

RSEM


This has been done in few cases:

i.e.: stress corrosion of inconel zones: in some case cracking occurs before expected: in the case of reactor vessel head penetration this has pushedto change the heads.

3213/Data collection and record keeping system3214/Maintenance histories, including:

All incidents are recorded in an incident log for follow up, analysis and information of Safety Authorities



- Information about function and equipment to be qualified are given to maintenance people of NPPs.

- In addition, qualified equipment maintenance sheets are being writen in order to point-out the particular point (a torque of a bolt, a radius of acable) which are sensitive for qualification tests and shall not be changed during normal maintenance interventions.

- Time to time checking of some functions are done: a large checking has been done before second ten year long outage.

278

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279

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EDF

DLE/01/07/00


(')��'RFXPHQWDWLRQ�511/ Self-assessment program512/ Peer reviews513/ Comprehensive reviews514/ Continuous improvement process

��5HYLHZ�7RSLFV�521/ Regulatory requirements for the AMP review None522/ Policy for the AMP review and improvement The program is improved by the PVD permanent activity. Reviews are decided time to time by high level management. Presentations are also

made on request to Safety Authorities (Specific Groupe Permanent).As EDF has developped and obtained the agreement of French Safety Authorities for a 10 years reassessment period, the results of ageingprograms are included in the overall checking launched prior to a ten year outage. At the moment such a checking has just been made for secondten years outage of 1300 PW units forecast (in 2004, head of the serie). But mainly no important decision in that matter will be made beforeVD3 (third ten year outage)

523/ Performance indicators524/ Updates of performance indicators

None

525/ Records of self-assessments526/ Plans and procedures of self-assessment527/ Capability of the NPP owner operator to evaluate

AMP effectiveness

Internal progress report, topic by topic are existing

528/ Records of peer reviews529/ Independence of the peer review teams5210/Results of the peer reviews5211/Effectiveness of corrective actions or

improvement programs

None

5212/Records of any comprehensive reviews Systematically done if any review5213/Roles of the operator and national regulatory

authorityAs per any problem the regulatory make decisions on proposal issued by the operator.

5214/Procedures of the comprehensive AMP review Not applicable

5215/Corrective actions arising Not applicable5216/A commitment of all members of the NPP staff to

continuous improvement5217/Relationship between the periodic safety review

program and any comprehensive AMP reviewThe reviews are performed in parallel to get the results at the same time, in order to be in position to implement any modification during thelong outage requested by regulation for hydraulic test of primary circuit.

281

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The initiatives of the nuclear power industry to improve life cycle management of nuclear powerplants, and those aimed at the safe and profitable extension of their design life, both share the needfor specific methodologies and must comply with the regulatory requirements established for thepurpose.

This has given rise to the development of guides and general documents defining suitablemethodologies for life management and for life extension.

However, the difference in the objectives pursued leads to differences in the methodologies. Indeed,it would seem unreasonable to demand full coincidence in terms of scope and requirements betweenthe licensing of life extension programmes and programmes aimed at ensuring safety and economicsoundness during design life.

Nevertheless, the technological basis of the methodologies, which is the knowledge of their state andof monitoring and control methods, features common factors that can be of use in the fulfilment ofthe different objectives.

�� 385326(

The purpose of this report is to compare the methodologies designed by NEI, IAEA and UNESA forlife extension and for ageing control of components that are important for safety and for lifemanagement, respectively.

This report examines and compares the requirements and methodologies defined in the followingdocuments:

• NEI (Nuclear Energy Institute) Guide 5- 10 Industry Guideline for Implementing theRequirements of 10 CFR Part -54 License Renewal Rule

• IAEA Technical Report series N° 338 Methodology for the Management of Ageing of NPPcomponents important to Safety

• UNESA Methodology for NPP Life Management (Metodologia para Gestion de Vida deUNESA)

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Following is a summary of the different methodologies in terms of objectives, requirements, structureand task sequence, as well as a comparison of their basic aspects to highlight existing differences.

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The NEI guideline, the regulatory requirement that gave rise to it (10 CFR Part 54, May 1995) andthe associated regulatory documents –the draft Standard Review Plan for License Renewal datedApril 2000, and Regulatory Guide DG-1047 dated August 1996–, will be issued as soon as the NEI

282

guideline be accepted, while waiting for completion of the revision process of current requests and ofother documents from Owners Groups.

The analysis included herein is based on the current issue of these documents, although the topicsthat are open for discussion are also indicated.

As a prime consideration, it should be noted that the purpose of application of this guideline is notlife management, nor the establishment of an ageing monitoring programme for the entire service lifeof the power plant. The only purpose of this document is to address compliance with the licenserenewal requirements indicated in 10CFR54 for a maximum period of twenty (20) years, based onand following a design life of forty (40) years maximum, which explains the strictness of some of therequirements.

This guideline is structured around the following steps of the License Renewal Rule:

a. Identification of the systems, structures and components (SSC) comprised in the scope andfunctions envisaged

b. Integrated Plant Assessment

1 Identification of the structures and components that will be subject to ageingassessment

2 Assessment of ageing effects

c. Time Limited Ageing Analyses (TLAAs)

d. License Renewal request documentation and contents

The most relevant aspects of implementation requirements and methods contained in Guideline NEI95-10 are briefly described and analysed below, for the purpose of comparison with the UNESA andIAEA methodologies.

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The Guideline defines the following:

a1. Safety-related SSC, as per 10CFR 50.49 (b) (1)

a2. Non safety-related SSC whose failure can impede items defined under al to perform theirfunction, including those that could cause seismic interaction with a 1 items

a3. According to safety analyses, all SSC that support functions covered by the following regulatoryrequirements:

• Fire protection

• Anticipated Transients Without Scram (ATWS)

• Environmental qualification

• Station Black-Out (SBO)

• Pressure Temperature Shock (PTS)

corresponding to 10CFR 50, 48, 49, 61, 62 and 63 respectively.

The functions considered in the scope selection are plant processes, conditions or actions that must befulfilled by the SSC in order to effect or support a safety function or required to comply with the fiveregulatory requirements listed above. If licensing bases include redundancy, diversity or defence-in-depth criteria, said criteria must be maintained in the identification of functions included in the scope.

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283

b1. Identification of structures and components that will be subject to ageing assessment

In this respect, the Guideline establishes that ageing assessment will be performed, on passive andlong-lived structures and components that perform or support the aforementioned functions

"Passive" applies to structures and components that cover the function without mobile parts orconfiguration changes; "long-lived" applies to SSC that are not subject to replacement beforeforty years, in accordance with a qualified life programme

b2. Assessment of ageing effects

The population selected in bl must be assessed to ensure that ageing effects, if any, are undercontrol so that their functions are assured; in their absence, improvements must be integrated inongoing programmes or specific programmes established for the purpose.

This analysis includes the study of the industry’s operating experience and of its applicability tothe plant.

The document establishes the following guidelines to demonstrate that existing monitoring andmaintenance practices are efficient to control ageing blow the limits established for compliance ofthe licensing bases during the life extension period

• Practices cover the structure or component analysed

• They would enable detection of ageing effect before loss of the structure or failure of thecomponent to perform its function

• They establish acceptance criteria and the obligation of appropriate corrective actions if suchcriteria are not respected

• Monitoring ensures reliable prediction with time for corrective actions

• Practices are submitted to sufficient administrative controls

If these guidelines are not fulfilled, then the practices require improvement or the integration ofnew ones.

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The Guide requires the identification and evaluation of these analyses to confirm that they are stillacceptable for the life extension period, either proceeding to a re-analysis for a sixty-year period withpositive result, or refining the analysis by eliminating conservatisms, or demonstrating by means ofageing monitoring that the evolution of ageing over time is not critical during the sixty-year period.

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The NEI guide 95-10 (rev. 1, January 2000) and the Standard Review Plan for License Renewal(April 2000) show discrepancies between NRC and Nuclear Industry. Once resolved, the NRC’ssupport to this NEI guide will be established in Regulatory Guide 1047 (currently in draft revisiondated August 1996). The open items are:

• Credit to programmes for inspection, condition monitoring and mitigation of existingdegradations, whose validity and efficiency need further justification, in the eyes of the NRC;while the industry argues that said justification, which is logical, should not lead to questioningthe current validity of these programmes and start together with the License Renewal a"relicensing" process equivalent to the process that took place prior to the current commercialoperation. The industry also argues that the justification of said programmes, which wereimposed by applicable regulatory requirements, should be small, if any, such as signalling itsconnection to the ageing control programme.

284

The GALL (Generic Ageing Lessons Learnt) report, issued in December 1999, will provide goodreferences to identify the credited programmes.

• As in the case of the programmes, it is the understanding of the industry that the acceptancecriteria employed in current programmes to ensure compliance with the licensing bases are alsovalid for life extension and require no further justification. In the case of new ageing controlprogrammes not required in the current licensing bases, said criteria should be established andjustified by means of what Guide NEI 95-10 defines as alarm values. This goes against theregulatory standpoint of justifying all criteria, including the ones included in the FSAR of thefacility.

• Controversy continues regarding the level (system or component) of the functions to bemaintained.

• The value of operating experience is debated, since the industry considers it as information ofoptional use in determining the efficiency of ageing control programmes, while the NRC makes itof mandatory use for the specific License Renewal programme.

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This study has been carried out by the IAEA within the framework of an international cooperationprogramme for better understanding of the degradation mechanisms that affect nuclear power plantsand for the development of efficient control methods. The study includes methods for the selection ofcomponents important for safety and whose ageing must be analysed. It also contains methods toanalyse said ageing effects, as well as the options for monitoring and mitigation.

The methodology defined in this study is oriented to ensuring safety during the service life of theplant by knowing, controlling and monitoring ageing mechanisms that could put to risk plant safety.

The methodology is structured as follows:

a. Selection process

1. Selection of the systems and structures that contribute to plant safety

2. Impact of component failure on system functions

3. Probability that ageing can cause component failure

4. Suitability of maintenance

b. Method for ageing assessment and control

Contents and development of the IAEA methodology are examined below, to, facilitate comparisonwith the ones defined in Guide NEI 95-10 and by UNESA.

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The selection process laid out in this method is carried out in two stages. The first corresponds to theselection of systems and structures that contribute to nuclear safety, while the second establisheswhich components will be submitted to ageing assessment.

To implement the selection stage, this method makes use of the safety classifications established ineach country and/or of the probabilistic safety analyses (if any).

The purpose of the second stage is to determine which components are more susceptible to ageingand insufficiently taken into account in the monitoring and maintenance programmes implemented inthe power plant. Component selection takes into account direct contribution to safety functions or therisk to impede such functions entailed by component failure.

285

Redundancy and diversity are not considered as exclusion criteria, since ageing is considered as acommon failure mode.

In the IAEA methodology, the preliminary selection of components, based on their contribution tosafety functions, is followed with the identification of the components among the list that aresubmitted to degradations that could cause their failure, in accordance with current knowledge andoperating experience of the plant and of the industry; it must also be established that thesecomponents are covered by inappropriate monitoring and maintenance practices with respect to thetype of ageing indicator, data acquisition techniques or delay in detection.

The conclusion of this selection based on simple criteria and on operating experience is theidentification of the component population that will be submitted to the systematic and stringentassessment that constitutes the second stage of the methodology, and which is described below.

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This task is performed on the population selected in item a) above. Its purpose is the detailed analysisof the components, of their characteristics, ageing mechanisms that affect them and gaps in themonitoring and mitigation practices implemented at the plant.

The first phase of this task is provisional, and its purpose is the identification of ageing effects, andthe proposal of immediate solutions, while establishing the areas requiring further inspection,analysis or research.

The second phase serves to confirm the conclusions reached in the first phase, or to detail theassessment and rectify the practices defined therein.

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The UNESA methodology comprises the following steps:

a. Selection of components, using life management criteriab. Examination of ageing mechanisms and selection of components subject to severe

degradationc. Evaluation of maintenance for life management and proposal of improvements

The most relevant aspects of the methodology as regards objectives, criteria, applicationrequirements and work guidelines are described below.

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It is carried out at two levels:• Selection of systems according to global criteria of safety, availability, replacement and cost• Selection of components according to seven different criteria covering nuclear safety, costs for

maintenance replacement or unavailability, and severe operating conditions

The criteria established in the Guide that are applied at the system level and those for componentscorrespond, in terms of nuclear safety, to the ones classified as safety -related in 10 CFR 50.49 (b) 1).As a rule in Spanish nuclear power plants, this classification also extends to systems that are notrelated to safety but whose failure may prevent or impede performance of safety functions.

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This assessment is performed on the component population selected in item a) above.

The assessment is based on the analysis of the construction and operating characteristics of eachcomponent or group of components, supported by the knowledge and systematic analysis of ageingeffects gained in the operating experience of the industry and in the specific experience of the plant,

286

as well as by what research programmes on ageing mechanisms (NPAR of the NRC, EPRI/DOE andUS Owners Groups, mainly) assign to these components or component groups. The assessment alsoidentifies the most suitable monitoring and/or mitigation methods, as direct reference for thesubsequent evaluation of maintenance.

The conclusions obtained are significant component/ageing mechanism pairs that will constitute thescope of maintenance evaluation.

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The purpose of this task is to evaluate the efficiency of maintenance practices for ageing controland/or monitoring, based on the inventory of practices and the systematic evaluation of the suitabilityof said practices in respect of each component/ageing mechanism pair identified.

Aspects to be evaluated are indicated in the following checklist:

• The degradation mechanisms identified in the component degradation data sheets are easilyidentified in the maintenance practice data sheets

• Maintenance actions and their frequency, as defined in plant programmes, are sufficient to detectageing and control proper operation of the component

• The subcomponents subject to degradation are properly listed

• Criteria or limits established for corrective actions are appropriate for the timely mitigation of thedegradation, thus preventing a potential loss of function

• The data required by the programme are sufficient to support the assessment methods

• All programmes are properly implemented and documented

• Maintenance efficiency has been proven

• The corresponding improvement proposals have been prepared

Non-compliance with one of the above items would lead to the preparation of the correspondingimprovement proposal.

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The following comparison outlines the common points (objectives and requirements) of the above-described methodologies, as well as the most significant differences among them. The analysis willfirst deal with the objectives covered by the methodologies, and follow with their comparison interms of contents, scope and requirements, in accordance with the sequence common to all threemethodologies, i.e.:

• Selection of scope• Ageing assessment• Evaluation of current maintenance and monitoring programmes and practices

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Guide NEI 95-10: the purpose of this guide is to interpret and develop on the requirements stated in10 CFR. Part 54 for license renewal for a period of twenty years maximum, upon expiry of thecurrent license (in the USA, usually corresponds to a forty-year design life).

The severity of the 10 CFR. Part 54 requirements is justified by the need to guarantee compliancewith the current licensing bases after design life has finished.

287

IAEA Methodology: its purpose is to define actuation requirements to ensure that nuclear safety willnot be affected during the plant service life (i.e.. design life) by ageing mechanisms in structures,systems and components.

UNESA Methodology: the objective pursued in this methodology is to establish requirements andtheir terms of application to ensure the safe and economic management of plant life.

Comparison of all three methodologies leads to the observation of the following differences in thebasic objectives they pursue:

Guide NEI 95- 10 Safety assurance against ageing effects during life extension period

IAEA Methodology Safety assurance against ageing effects throughout service life

UNESA Methodology Safe and economic operation during service life

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Table 3.4-1 lists the basic criteria and requirements of the selection process in each methodology.

The main differences in the selection process of the three methodologies are summarised below:

Guide NEI 95-10 Initial selection includes all SSC that are safety-related and those thatare not safety-related and whose failure can prevent performance ofsafety functions, in addition to those supporting compliance withcomplementary regulatory requirements (see Table 3.4-1). However,ageing assessment is applied only to passive and long-lived structuresand components

IAEA Methodology Initial selection includes all SSC that are safety-related and those thatare not safety-related and whose failure can prevent performance ofsafety functions. However, strict and systematic ageing assessment isonly applied to the population which, exclusively in accordance to theexperience gained by the industry and to the plant specific experience,are subject to failure by ageing and are not covered by appropriatemaintenance for ageing mitigation and/or monitoring

UNESA Methodology Initial selection includes safety-related SSC (see section 3.4.a) as wellas those that have significant impact on availability, replacement andcost.

The entire population selected is submitted to systematic ageingassessment; when ageing effects are severe, maintenance efficiency isassessed

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Table 3.4-2 compares the basic requirements of the ageing assessment process in each of the threemethodologies.

The analysis of these differences is carried out in two steps: in the first, the assessment of potentialageing and its effect on the components and their functions is studied, while the second examines theassessment criteria for maintenance practices and programmes, and their suitability for ageing controland/or monitoring.

288

The differences among the three methodologies are summarised below:

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The purpose of ageing assessments is to detect ageing mechanisms that could prevent safety functionperformance in the course of the life extension period.

The assessments are aimed at determining the severe effects of ageing on components and structures,and at confirming maintenance efficiency to mitigate and/or monitor said effects, keeping thecomponents in conditions to fulfil the functions imposed in the licensing bases during the lifeextension period.

Lack of sufficient information to accurately assess ageing effects and their evolution requires theestablishment of a specific inspection programme for license renewal.

The identification and revalidation of the Time Limited Ageing Assessments (TLAAs) are alsomandatory for the life extension period.

Maintenance evaluation is performed with criteria similar to the ones in the other methodologies, butfor the life extension scenario.

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The purpose of ageing assessments is to detect ageing mechanisms that could prevent safety functionperformance in the course of the plant service life period

The assessments are aimed at determining the severe effects of ageing on components and structures,and at confirming maintenance efficiency to mitigate and/or monitor said effects, keepingcomponents within safety margins for them to fulfil their functions during service life.

Lack of sufficient information or of precise knowledge of degradation mechanisms that generateuncertainties in the initial "provisional" assessment (phase I) of ageing effects require detailedanalyses (phase II) to clear said uncertainties.

Maintenance evaluation is performed with criteria similar to the ones in the other methodologies, butfor the service life scenario.

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The purpose of ageing assessments is to detect ageing mechanisms that could prevent performance ofthe functions necessary for safe and economic operation during plant service life.

The assessments are aimed at determining ageing effects on components and structures, and atidentifying the ones that are severely affected and therefore require a detailed analysis ofmaintenance practices to confirm ageing control during service life.

The entire process is summarised in the component degradation data sheets and in the maintenancepractice data sheets that are prepared to facilitate the systematic evaluation of existing practices withthe subsequent improvement proposal.

Maintenance evaluation is performed with criteria similar to the ones in the other methodologies, butfor the service life scenario.

289

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SSC selection criteria − Safety-related,

− Non safety-related. whose failurecould prevent or impedeperformance of safety functions(includes seismic interactions)

− SSC supporting compliance withspecific regulatory requirements(PC, EQ, PTS, ATWS and SBO)

− Safety-related (classification ofeach country)

− Non safety-related. whose failurecould prevent or impedeperformance of safety functions(includes seismic interactions)

It is recommended to supplement: thedeterministic selection withprobabilistic analyses, if any.

− Systems classified as safety-related

− Systems with significant impact onavailability, replacement and cost

− Components according to criteria ofsafety, maintenance, replacement orunavailability cons and severeoperating conditions

Additional selection criteria Redundancy and diversity are notconsidered exclusive

Redundancy and diversity are notconsidered exclusive

Redundancy and diversity are notconsidered exclusive

Selection of structures andcomponents for ageingassessment

Only passive and long-lived structuresand components are submitted toageing assessment

Systematic ageing assessment applied.only to components subject to failureby ageing and not covered by efficientmaintenance according to operatingexperience of the plant and of theindustry

All components selected are submittedto ageing assessment.

Only severely affected components aresubmitted to maintenance evaluation.

290

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5HTXLUHPHQW 1(,�� ,$($�0HWKRGRORJ\ 81(6$�0HWKRGRORJ\Analysis of ageingmechanisms(purpose)

Analyse ageing mechanisms to,determine which could impede orprevent safety function performanceduring life extension period

Analyse ageing mechanisms todetermine which could impede orprevent safety function performanceduring service life

Analyse ageing mechanisms todetermine which could impede orprevent performance of the functionsnecessary for safe and economicoperation during service life

Analysis of maintenanceefficiency (purpose)

Demonstrate whether ageing effectsare efficiently controlled and/ormonitored to ensure functionperformance within licensing basesduring life extension period

Demonstrate whether ageing effects,are efficiently controlled and/ormonitored to ensure functionperformance within licensing basesduring service life

Demonstrate whether ageing effects areefficiently controlled and/or monitoredto ensure performance of the functionsnecessary for safe and economicoperation during service life

Contents and detail level ofageing assessments

Review and evaluate availableinformation regarding design,materials, operating conditions,performance, operation andmaintenance history and informationregarding operating experience andinternational research on degradationmechanisms affecting each component

Review and evaluate availableinformation regarding design,materials, operating conditions,performance, operation andmaintenance history and informationregarding operating experience andinternational research on degradationmechanisms affecting each component

Review and evaluate availableinformation regarding design, materials,operating conditions, performance,operation and maintenance history andinformation regarding operatingexperience and international researchon degradation mechanisms affectingeach component

Lack of such information makes itnecessary to use exclusivelyinformation based on industryexperience, trend calculations orhistorical analyses; in case these proveto be insufficient, an inspectionprogramme is additionally required.Identification and revalidation for thelife extension period of applicableTime Limited Ageing Assessments(TLAAs) are also mandatory

Such assessments, which areconsidered provisional, mustsystematically take all of the above inconsideration for each component orstructure, and conclude with theidentification of severe ageing effects,location of incidence and the necessaryparameters for follow-up

Such assessments, which are consideredprovisional, must systematically take allof the above in consideration for eachcomponent or structure, and concludewith the identification of severe ageingeffects, location of incidence and thenecessary parameters for follow-up

291

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As an alternative to these specific assessments, theGuide admits other options based on strictjustification:

• Analysis based on applicable references (eg,Technical Reports from Owners Groups)

• Through condition and/or performancemonitoring programmes

Uncertainties in the knowledge andgrowth of ageing mechanisms should becompleted by more detailed analysesconfirming provisional assessments andallowing better definition of trends andmonitoring parameters

The assessments include identificationof the mitigation and monitoringmethods in accordance with industryexperience. All of the above is gatheredin Component Degradation Sheets tofacilitate the maintenance evaluationprocess

Evaluation criteria formaintenance practices

List of necessary attributes to prove the suitability ofthe practices:

• The component or structure is included in themaintenance programme or instruction analysed

• Ageing effects are detected before failure

• The practice includes acceptance criteria andensures the implementation of corrective actionswhose extent is sufficient in scope and in time

• Trend monitoring makes appropriate predictionpossible

• The practice is placed under appropriateadministrative controls

Aspects evaluated are:

Monitoring

Suitability of parameters

Accuracy and reliability of dataacquisition techniques and analysisof parameters

Provenness of ageing assessmenttechniques

Mitigation.

Analysis of the efficiency of thepractices in mitigating degradationso that performance of the functionsis not at risk

The checklist applied to assessment is:

Degradation mechanisms identified inComponent Degradation Sheets areclearly identified in the MaintenancePractice Data Sheets.

The actions and action frequenciesestablished in plant programmes aresufficient to detect ageing effects andcontrol proper component operation

Subcomponents that are subject todegradation are listed

The criteria and limits established forthe implementation of a correctiveaction ensure timely mitigation of thedegradation mechanism that could leadto a possible loss of function

292

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The data required in the programme are sufficient tosupport assessment methods

All programmes are correctly implemented anddocumented

Maintenance efficiency has been demonstrated

Corresponding improvement proposals have beenprepared

The characteristics and scope of maintenancepractices, their frequency and acceptance criteria aresummarised in the Maintenance Practice Data Sheetsto facilitate this evaluation

293

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The methodologies defined in the Guide NEI 95-10 and in the IAEA’s Technical Report presentcontents and requirements that are shallow, preventing the detailed analysis or comprehensivecomparison of some aspects. This explains why the comparison was established at a levelpermitting definition of the requirements and application examples of both documents, avoidingrisky interpretations. In the case of Guide NEI 95-10, as mentioned before, there is an open debatebetween the industry and the NRC regarding some of the requirements and accuracy with respect to10 CFR Part. 54. The following basic conclusions were obtained based on the above, and asdescribed in Chapter 3 of this report:

1. In respect of the objectives, while the UNESA methodology is designed for the safe andeconomic management of plant service life, the IAEA methodology confines itself to safetyassurance against ageing during service life and the Guide NEI 95-10 to license renewal,assuring safety during the life extension period

2. As regards the scope of the respective programmes, there are no significant differences (seeTable 3.4- 1), although the UNESA methodology covers the components and structures withsignificant impact on availability, complexity of replacement, ALARA effects, and cost of repairor replacement.

It should be noted that, within the Selection chapter, the UNESA methodology includes thesystematic ageing assessment of all components and structures selected, while the IAEAmethodology limits said assessment to the components; and structures pre-selected fromindustry and plant experience, and the Guide 95-10 restricts ageing assessment to the selectedcomponents and structures that are passive and long-lived.

3. With respect to ageing assessment, application examples seem to indicate a certain similarityin their level of detail (with the scope exceptions indicated above), although Guide NEI 95-10 ismore comprehensive and rigorous in terms of documentation and presentation. This Guide alsorequires compensating lacks of information for variables that are representative of ageingeffects, and of precise knowledge of their evolution, with a specific inspection programme priorto license renewal. This requirement, which seems sensible for license renewal, is not justifiedfor life management or ageing monitoring programmes during the plant service life, since theevolution of ageing effects is controlled so that additional inspection and monitoring measuresare intensified as required.

Finally, the evaluation of the suitability of maintenance practices and programmes with respectto the monitoring and control of ageing effects is approached follows very similar steps in allthree methodologies (see Section 3.4). Again, the scope exceptions indicated above must betaken into account, as well as the fact that in the case of Guide NEI 95-10 they are limited topassive components, giving credit to the programmes and requirements; (such as theMaintenance Rule) applicable to active components.

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The documents used as references for this report are listed below.

5.1 UNESA, “Guía de Selección de Componentes”. Documento nº GVR-EA-GU-00101.

5.2 UNESA, “Guía de Evaluación de Prácticas de Mantenimiento”. Documento nº GVR-EA-GU-04101.

5.3 UNESA, “Estudios de Mecanismos de Degradación de las Centrales Piloto”. Documentos nºGVR-EA/IN-IT-03301 a 03317.

294

5.4 UNESA, “Evaluación de Prácticas de Mantenimiento de las Centrales Piloto”. Documentosnº GVR-EA-IT-04401 and 04402.

5.5 International Atomic Energy Agency, “Methodology for Ageing Control in ComponentsImportant for Nuclear Safety”. Technical Report Series Nº 338.

5.6 Nuclear Energy Institute, “Industry Guideline for Implementing the Requirements of 10CFR Part 54”. NEI 95-10 Rev. 1, January 2000.

5.7 US NRC, “Working draft Standard Review Plan for License Renewal”. WD-SRP-LR, April2000.

5.8 US NRC, “Draft Regulatory Guide DG-1047”. August 1996.

295

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This appendix includes a bibliography listing the various documents identified in the course of thepresent study. They are classified in two groups:

− Reports, classified according to author or organisation, with title, date, number of pages, reportreference and status mentioned. Where known by the authors, the status of the documents ismentioned,

− Publications, classified according to their first author, with the company, title, conference orjournal, page, date and place mentioned.

The identification number corresponds to the identification given in the list of Section 8, which islimited to the references explicitly referred to in the main text of this report.

A short summary is given for a selected number of papers or reports. Where reports have beendeveloped in Appendix 2, the information in this column is limited to a reference to the appropriatesheet.

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Reports: Pages 296 – 330,

Publications: Pages 331 – 362.

296

Appendix_6: Rapports

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6WDWXV 6KRUW�VXPPDU\

Nuclear Power plant Common AgingTerminology - Final Report

EPRINuclear EngineeringInstitute

01/11/1992 EPRI TR-100844

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Canada

Expert Group onPlant LifeManagementMeeting(OECD/NEA/PLIM)

01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Czech & Slovak


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Finland


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - France


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Germany


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - India


01/04/1992

297

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Korea


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Netherlands


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Spain


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Sweden


01/04/1992

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - US


01/04/1992

AEOD Engineering EvaluationReport - A Review of Water HammerEvents after 1985

OECD/NEA 01/02/1991 AEOD/E91-01

Completion of the Fatigue ActionPlan

US NRC SECY-95-245

Cracking in Feed Water pipingsystem

US NRC 16/10/1979 NRC-IE BulletinN°79-13

Pressuriser surge line thermalstratification

US NRC 20/12/1988 NRC-BulletinN°88-11

Service Water System Problemsaffecting safety-related equipment(July 18, 1989) + Supplement 1

US NRC 04/04/1990 NRC-GenericLetter 89-13

Thermal cracking of Feed Waterpiping to steam generators

US NRC 24/03/1993 NRC-InformationNotice N°93-20

Thermal stratification in Feed Waterpiping system


298

Thermal stress in piping connectedto reactor coolant system (June 22,1988) + Supplement 1 (June 24,1988) + Supplement 2 (August 4,1988) + Supplement 3

US NRC 11/04/1989 NRC-BulletinN°88-08

Unexpected piping movementattributed to thermal stratification


Aging Management Evaluation forClass 1 Piping and AssociatedPressure Boundary Components

WOG Report 01/08/1996 WestinghouseReport

Aging Management Evaluation forClass 2,3, and Nonsafety ClassValve Bodies, Pump Casings, Pipingand Supports, and Ductwork


Aging Management Evaluation forElectrical Distribution, Control, andMonitoring Equipment


Aging Management Evaluation forPressurized Water ReactorContainment Structure


Aging Management Evaluation forSeismic Category 1 Structures


License Renewal Evaluation: AgingManagement for Reactor CoolantSystem Supports


Abbott S.L. , B.A.Bishop, S.L.Anderson, M.Blaszkiewicz, E.Blocher, C. Child, D.Kurek, T.R. Mager

Aging management evaluation forreactor pressure vessel


AMES SteeringCommittee

Important Items of Ageing Research AMES Report 20 November1998

Rev.1

Atkinson I. Continuous on-line monitoring ofNPPs components

AEA Technology,EuropeanCommission

01/12/1995 EUR 18333 EN

Bedzikian G. Tenue en service des cuves destranches REP 900 Mwe. Dossier desynthèse

EDF 01/10/1997 D4002-42-10/97.0676

299

BenchmarkQualification Group(qualification ofelectro-mechanicalequipment)

Guideline for the evaluation ofEuropean practices on the harshenvironment qualification ofelectrical and I&C equipment

Nuclear science andtechnology,EuropeanCommission

1998 EUR 17563 EN

BenchmarkQualification Group(qualification ofelectrome-chanicalequipment)

A comparison of European practicesfor the qualification of electrical andI&C equipment important to safetyfor LWR nuclear power plant


1996 EUR 16246 EN

Berg R. , J. Shao, G.Krencicki, R. Giachetti

Aging Management Guideline forCommercial Nuclear Power Plants -Stationary Batteries

SANDIALaboratories

mars-94 SAND93-7071

Berg R., M. Stroinski,R. Giachetti

Aging Management Guideline forCommercial Nuclear Power Plants -Battery Chargers, Inverters andUninterruptable Power Supplies

SANDIALaboratories

Feb 94 SAND93-7046 Finalreport

Bolvin M. Projet Durée de vie - Tours desaéroréfrigérants principaux - Rapportde synthèse

EDF avr-92 Ref EDF:F92SE1338

Booker S. , D. Katz,N. Daavettila, D.Lehnert

Aging management guideline forcommercial nuclear power plants-pumps

SANDIALaboratories

mars-94 SAND-93-7045

Carlson R. W. Aging Management Evaluation ofthe residual heat removal system


Carlson R.W. Aging Management Evaluation ofthe Residual Heat Removal Systemfor Westinghouse PWRs

EPRI mars-96 TR-105135 FinalReport

Chauvel D., M. Bolvin Durée de vie des tours deréfrigérants

EDF 31/03/1992 F92SE1018

Chevet P.F. Projet "Durée de vie" des REP DSIN sept-93 Ref. EDF:F93SE2049

Churier-Bossenec H. Vieillissement des produits moulésdu circuit primaire principal. Coudeschauds des tranches 900 Mwe.Synthèse des études

EDF/Septen 02/09/1996 E-N-M-RE-96,1054

Internalreport(restricted)

The reports presents the justification of the in-service acceptance of cast elbows used in the hotlegs of the main primary system of French 900 MweNPPs, for a 40 years operation. It presents thephysical aging process, the previsions of toughnessreduction, and the fast fracture analyses made todemonstrate the acceptability of cast elbows. Aleak-before-break evaluation is included as acomplement.

300

Clark R.L. Effects of aging and service wear onmain steam isolation valves andvalve operators

ORNL mars-96 NUREG CR-6246; ORNL-6814

CO/CP Operating feedback of EDF’s coolingtowers: civil work and equipment

EDF 31/08/1989 F89RT0035

Conklin J.E. Aging Management Evaluation forHeat Exchangers


Conte M., G. Deletre,J.Y. Henry

Approche des problèmes de sûretéliés au vieillissement descomposants des ecntralesnucléaires

CEA 30/08/1988 Ref. EDF:E1989E130512

Davies L.M. , A.D.Boothroyd, L. Ianko

Aspects of Plant Life Assurance andPlant Life Management

IAEA Report Int. Conf. Onthe NuclearPowerOption,Vienna, 5-9Sep. 94

Paper IAEA-CN-59/40

De Bauw K. Arbre de décision pour les alarmesvibratoires des turbines nucléaires

Laborelec Report 30/03/1998 CONDE-MON-97-0065F

Doroshuk B.W., B.M.Tilden, D.R. Hostetler,C.A. Negrin

Nuclear Plant Life CycleManagement Information Systems

EPRI déc-94 TR-103858 FinalReport

Doroshuk B.W., B.M.Tilden, D.R. Hostetler,D.J. Klein, C.A.Negrin

Calvert Cliffs NPP Life CycleManagement License RenewalProgram: System, Structure andComponent Screening

EPRI Report September1994

TR-103158 This report describes and documents the"screening method" and procedures to determinethose systems, structures and components withinthe scope of licence renewal for the Calvert CliffsNPP License Renewal Application. This work wasperformed based on the initial Licence Renewalpublished in 1991, however the results are of valuesince scoping is similar in many aspects betweenthe initial and current license renewal rules.

Doroshuk B.W., G.R.Doughty, S.J.Marmaroff

Calvert Cliffs NPP Life CycleManagement/License RenewalProgram: Nuclear Plant AssetManagement Case Study

EPRI sept-95 TR-104615 FinalReport

Doroshuk B.W., M.E.Bowman, S.A. Hardin,M.D. Lusby

Calvert Cliffs NPP Life CycleManagement / License RenewalProgram: RPV Evaluation


301

Dorosshuk B.W., G.R.Doughty

Calvert Cliffs NPP Life CycleManagement/License RenewalProgram: Steam generator decisionanalysis case study


Dragunov Y.G. Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Russia

OKBExpert Group onPlant LifeManagementMeeting(OECD/NEA/PLIM)

01/04/1992

Feinroth H., R.S.Walker, C.A. Negin,M.A. Rowden, J.R.Kraemer

Regulatory Considerations forextending the Life of Nuclear PowerPlants

Grove Engineering,National Environm.Studies Project ofthe Atomic Ind.Forum

01/12/1986 AIF/NESP-040

Forsyth D.R., B.L.Silverblatt, T.R.Mager, W.A. Bamford,J.A. Tortorice, J.T.Crane, I.L.W. Wilson

Aging Management Evaluation forReactor Internals


Fournier I. Recherche des défaillances sur lecircuit primaire des tranchesaméricaines

EDF-DER Report 25-avr-90 HP-16/90,16 In the context of the Sysfac project, the reportpresents the synthesis of the localisation and originof fatigue degradations observed on US PWRNPPs.

Frederick G., P.Hernalsteen

Belgian approach to steamgenerator tube plugging for primarywater stress corrosion cracking

Belgatom (TEE/LB) 01/03/1999 EPRI NP-6626-SD

Fresco A., M.Subudhi, W. Gunther,E. Grove, J. Taylor,S.K. Aggarwal

Managing aging in nuclear powerplants: insights from NRCmaintenance team inspectionreports

NRC BrookhavenNat. Lab.

déc-93 NUREG CR-5812/ BNL-NUREG52309

Frog members: P.Namy

16th steering committee minutes ofmeeting

FROG, TEE Report 29/10/1998 CNT-KCD/6F/154413

Goossen J.E., J.C.Matarazzo, R.J.Maceyak, C.H. Boyd

Baffle-Former Bolt Program for theWestinghouse Owners Group -Phase 1: Plant categorization


Guerra D. Aéroréfrigérant - Etude économiquesur la maintenance du corpsd'échange.

EDF 15/04/1994 F94RT0344

302

Guillas J. Vieillissement des condensateurselectrolytiques hors tension et soustension. Elements techniques pourl'évolution de la doctrine demaintenance des centralesnucléaires

EDF juil-96 Ref EDF:E1996H502253

Restricted

Hasemian H.M., D.W.Mitchell, R.E. Fain,K.M. Petersen

Long term performance and agingcharacteristics of nuclear plantpressure transmitters

NRC mars-93 Ref. EDF:E1995E100154

Hassan M., S.Uryas'ev, W.E. Vesely

Sensitivity and uncertainty analysesin aging risk-based prioritizations

BrookhavenNational Lab.

1993 BNL-NUREG-49610

Heep W., P. Vögtlin,P. Ami, D. Steudler

Guide Manual for MaintenanceCertification of StructuralEngineering

Group of SwissNuclear Power PlantOperators

31/10/1997 The engineering structures classified according toGuidelines R04 of the Swiss Nuclear Inspectoratein BK1 are considered. The engineering structuresconsidered are established in the safety report. Themaintenance plan addresses structures andcomponents which can decisively impair thefunctions relevant to safety as a result of the agingprocess of technical material.

Huet M. Essais de vieillissement prolongésur une baie ébulliomètre et deuxafficheurs

CEFSE janv-96 Ref. EDF:F97SE3065

Rapportd'essais

Huet M. Essais d'irradiation de vieillissementsur des liaisons coaxiales deschaînes neutroniques intermédiairesCNI

NMEL 07/04/1995,20-01-1995

Ref. EDF:F95SE1840,F95SE1838

Ithurralge G. Projet "Durée de vie" - Rapport deconstat sur le puits de cuve - REP.

EDF juil-93 Ref EDF:F89SE1480

Kaushansky M.M.,K.R. Balkey, B.A.Bishop, T.A. Meyer

The nuclear power plant agingmanagement process POWER-GENEurope'93. Volume 6: nuclear powerplant operations and maintenance

5/POWERGEN-C janv-93 Ref. EDF:E1994H200730

Kisisel I.T., R.L. Kurtz,J. Sinnappan

Plant systems / components agingmanagement - 1993

1993 PVP Conf. 25-29 Jul1993

CONF-9307/02-Vol.252

Proceedings

Kisisel I.T., R.L. Kurtz,J. Sinnappan, T.V.Narayanan

Plant systems/ components agingmanagement 1994

1994 press. Vess.And piping Conf.

1994 CONF-940613

Klanica F., C. Gay License Renewal Evaluation: AgingManagement for Class 1 Piping andAssociated Pressure BoundaryComponents


303

Lacoste A.C. REP - Programme "durée de vie" DSIN 18/10/1996 Ref. EDF:F96SE2658

Langlois R. EDF R2C project: results of thestudies related to the observation ofthe behavior and the ageing of theprocess control equipment installedin EDF 900 Mwe PWR units

EDF 01/08/1995 EPRI CongressRef. EDF:E1995E100810

Lapay W.S., R.E.Funkhouser, C.C.Kim, R.T. Jozwiak

Aging Management Evaluation forsafety Class 1 Piping Supports


Lapay W.S.; C.Y.Yang

Aging Management Evaluation forReactor Coolant System Supports


Lapsay W.S., C.Y.Yang

Aging Management Evaluation ofReactor Coolant System Supportsfor Westinghouse PWRs


Le Brun A., F. Billy Non-destructive fatigue damageassessment using ultrasonic andmagnetic measurement on metallicmaterials

EDF Report October 1993 94NB00056,ISSN 1161-0611

The paper deals with indicators of damage ofmetallic structures based on non-destructivemeasuring methods. Degradation mechanisms arebasically due to mechanical fatigue damage,thermal aging, radiation embrittlement andcorrosion. The paper exposes the results obtainedusing two types of techniques: magnetic andacoustic, with a view to assessing damage bymechanical fatigue.

Lee B.S. The effects of aging on BWR coreisolation cooling systems

BrookhavenNational Lab.

oct-94 NUREG CR-6087/ BNL-NUREG52390

Lee T.R. Public perception on ageing plants févr-92 Ref. EDF:E1992E130094

Levy D.B. , C. Feltin Approche conceptuelle desproblèmes de sûreté posés par levieillissement des centralesnucléaires

IAEA Congress 01/01/1988 Ref. EDF:E1988H548695

Liu W.C. , P.T. Kuo,S.S. Lee

Aging management of nuclear powerplant contaiments for license reneval

NRC sept-97 NUREG 1611

Meyer L.C. Nuclear Plant Aging research onhigh pressure injection systems

NRC août-89 Ref. EDF:E1990E150065

Moinereau D., J. Cl.Masson

Comportement en Fatigue thermiquede composants de centralesnucléaires

Collection de notesinternes de la DER,EDF

13/06/1905 ISSN 1161-0611

304

Monteiro P.J.M., J.P.Moehle

Stiffness of reinforced concrete wallsresisting in-place shear. Tier 2:Aging and durability of concreteused in NPPs.

EPRI déc-95 Ref. EDF:E1997E100329

FinalReport

Moyers J.C. Aging of control and service aircompressors and dryers used innuclear power plants

ORNLNRC

juil-90 NUREG/CR-5519-Vol.1ORNL-6607-Vol.1

Nad L. Nuclear safety - Diagnosis, safetyand aging of concrete structures inNPPs - Thermal properties andmechanical behaviour of the Penlyhigh resistance concrete

NTMS juin-95 Ref. EDF:F97SE2860

Namy P. RHRS elbow cracking at Civaux 1 Framatome 11/08/1998 FRA OG 536Narayanan T.V. Criteria for Approving Equipment for

Continued OperationWRC Bulletin WRC Bulletin 380 145 reports were reviewed and various experts

consulted. The report proposes conclusions andrecommendations for ASME Code-related activitiesrelated to continued operation methods and criteria.

Naus D.J. , C.B.Oland, B. Ellingwood,Y. Mori, E.G. Arndt

Aging of concrete containmentstructures in nuclear power plants

ORNL 1992 CONF-920541-2

Naus D.J., C.B.Oland, B.R.Ellingwood

Report on aging of nuclear powerplant reinforced structures

NRCJohns Hopkins Univ.

mars-96 NUREG:CR-6424

Newton R.A., T.A.Meyer

Life Cycle Management - LicenseRenewal Program - Program Plan

WOG Report 01/05/1996

Nichols R.W. A state of-the-art review ofcontinuous monitoring andsurveillance techniques in relation toreactor pressure circuit integrity

EC 1991 EUR-13409

Novak S.,M. Podest

Safety aspects of the ageing andmaintenance of NPP (Symposiumoverview)

IAEA Report janv-87 Ref EDF:E1988H547233

Pachner J. NPP ageing and life extension safetyperspectives

févr-92 Ref EDF:E1992E130091

Persoz M. Statistical models for thermal ageingof steel materials in NPP

EDF mai-96 Ref. EDF:E1996H400932

305

Ramirez I.M., M.T.Aguado Esteban, L.T.Roa, D. Foster, M.Sladovic

Preparatory work for an indicativeprogramme related to ageing issuesto be handled by the WGCS

Tecnatom December1997

Study contractB4-3070/96/000295/MAR/C2

FinalReport

The aim of the report is to collect the work onageing conducted by the internationalorganisations, research institutions, utilities, etc.,identifying and evaluating the objectives, progressand results. The programs related with ageing ofLWRs are analysed, detecting the status and themonitoring techniques and ISI activities performedto support continuous operation of components, aswell as identifying the technical issues that needR&D effort and work on their resolution.

Rotival Analyse de l’influence, sur levieillissement des tranches, desmodes d’exploitation et de gestion

EDF 27/08/1987 F87SE1176

Rousseau M. REP de 1300 et 1450 Mwe -Enceintes de confinement

DSIN 09/07/1998 Ref. EDF:F98SE2640

Schinazi C. Rapport de réunion relatif à laprésentation des résultatsd'inspection US d'une barrièrethermique de pompe primaire à Doel3

TEE Report 08/02/1999 KCD/4NT/20317/00

Schinazi C., J.M.Cherasse

Fissuration des barrières thermiqueset autres composants des pompesprimaires.Note d'information sur l'état de laproblématique en France etproposition d'action pour lescentrales belges.

TEE Report 07/04/1997 CNT-KCD/4NT/7450/00

Schulz H. CSNI, PWG-3 Technical PositionDocument. Plant AgeingManagement. Providing a technicalbasis for long-term operation of lightwater reactors.

CSNI May 30, 2000

Secretariat ofOECD/NEA

Policy and effective management ofnuclear power plant lifemanagement

OECD/NEA Report 15 April, 1999 Firstdraft,version 5

General Framework.

Shah V.N., A.G.Ware, A.M. Porter

Review of industry efforts to managepressurized water reactor feed waternozzle, piping and feed ring crackingand wall thinning

Idaho NationalEngineeringLaboratory

01/03/1997 NUREG/CR-6459

306

Shah V.N., S.K.Smith, U.P. Sinda

Insights for Aging management oflight water reactor components:Metal Containment

Idaho Nat. Eng. Lab. mars-94 NUREG/CR-5314-V5

Techn.Report

Simola K. The role of reliability methods inNPP ageing management

janv-98 Ref. EDF:E1999E100177

Stirzel C.W., J.E.Conklin

Aging management Evaluation forPressure Vessels


Suzuki M., H. Arai, T.Hidaka

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension - Japan

JAERI, JAPEIC,Expert Group onPlant LifeManagementMeeting(OECD/NEA/PLIM)

01/04/1992

Syvester R.L., M.A.Gray

Aging management Evaluation forPressurizers


Toman G., R.Gazdzinski, E.O’Hearn

Aging Management Guideline forCommercial Nuclear Power Plants -Motor Control Centers

SANDIALaboratories

Feb 94 SAND93-7069 Finalreport

Toman G., R.Gazdzinski, K.Schuler

Aging Management Guideline forCommercial Nuclear Power Plants -Electrical Switchgear

SANDIALaboratories

juil-93 SAND93-7027

Valibus L. French perspective on lifemanagement of NPP

EDF-Septen 20/03/1996

Vandenbussche R. Réévaluation de la durée de viequalifiée des indicateurs de positionCROSBY

TEE Report 01/06/1994 KCD4/4NT/7770/01

Vesely W.E., R.E.Kurth, S.M. Scalzo

Evaluations of core melt frequencyeffects due to component aging andmaintenance

NRC juin-90 NUREG/CR-5510SAIOC-89/1744

Vora V.V., D.E.Prager, D.C.Bhowmick, K.R. Hsu

Integrity Evaluation of the PrimaryLoop piping including the effects ofthermal aging using LBB technologyfor a typical EDF 3 loop NPP

Westinghouse déc-95 EDF Cl. Nr:F96SE0725

Wagner-Rousseau D. Mécanismes de vieillissement etd'endommagement dans quelquesalliages pour application nucléaire

Thèse CNRS /Ecole Centrale

15-janv-99

WGCS Compte-rendu de la réunion duWorking Group Codes andStandards

EC - DG XI 30/06/1998 CNT-KCD/4AR/8525/00

Whiteman G.W. Aging Management Evaluation forSteam Generators


307

Wolford A.J., C.L.Atwood, W.S.Roesener

Aging data analysis and riskassessment: development anddemonstration study

Idaho Nat. Eng. Lab. Aug 92 NUREG/CR-5378EGG-2567

Techn.Report

Assessment of AgingCountermeasures of Electric UtilityCompa-nies for Nuclear PowerPlants and Future Appro-aches toManagement of Plant Aging

Agency of Naturalresources andEnergy Ministry ofInternational Tradeand Industry

Feb. 1999 The paper describes the key points of the Report on"Basic Policy on Aged Nuclear Power Plants"issued by the MITI in Japan. The assessment ofreports provided by Electric Utility Companies isaddressed and future approaches to managementof plant aging are presented including:comprehensive facility management system,development of technical codes/standards inresponse to aging, and promotion of technologydevelopment.

Survey of existing, planned andrequired standards

AMES Report 1995 CD-NA-16313 EN Public The objective of the report is to provide a survey ofnational standards used in European countries toperform and to evaluate test results in order todescribe degradation of reactor pressure vesselsteels and welds of light water reactors duringservice.

Risk-Based Inservice Testing:Development of Guidelines

ASME ResearchReport

1996 CRTD-Vol. 40-2 Public This document is the second of a series, which isbeing developed by a multidisciplinary ASMEresearch task force. Vol.2 is the first specificapplication directed at the inservice testing of LWRnuclear power plant components, particularly forpumps and valves.

A REVIEW OF FORMULAS FORPREDICTING IRRADATONEMBRITTLEMENT OF REACTORVESSEL MATERIALS

CEA nov-96 CD-NA-16455-EN-C

EFFECT OF IRRADATION ONWATER REACTOS INTERNALS

CEA/TECNATOM/VTT

juin-97 CD-NA-17694-EN-C

Common ageing terminology Common NEA, CECand IAEA report

July 1999 A glossary useful for understanding and managingthe ageing of Nuclear Power Plant systems,structures and components.

Revisiones Periódicas de laSeguridad de centrales nucleares

CSN 1996 GS-1.10

A Risk-Informed Flaw ToleranceApproach for Increasing ASMESection XI, App. G PT Limits

EPRI juil-97 TR-107451 FinalReport

Cable Aging Management Programfor DC Cook NPP Unit 1 and 2

EPRI mai-97 TR-106687 FinalReport

308

Calvert Cliffs NPP Life CycleManagement/License RenewalProgram:System, Structure andComponent Screening

EPRI janv-95 TR-103158 FinalReport

Calvert Cliffs Nuclear plant LicenseRenewal Application Technical basis

EPRI janv-98 TR-106843 InterimReport

Calvert Cliffs Nuclear Power PlantLicense Renewal Application

EPRI janv-98 TR-11031-CD

Condition Monitoring Program for4kV Environmentally QualifiedMotors


Crack Growth and MicrostructuralCharacterization of Allow 600 HeadPenetration Materials

EPRI mai-96 TR-105958 InterimReport

Demonstration of the Managementof Aging Effects for B&WPressurizers

EPRI janv-98 TR-106106

Demonstration of the Managementof Aging Effects for Reactor CoolantSystem Piping


Effect of NPP, Thermal Stability, andCorrosion Properties of Allow 718and 718-Based Superalloys


End-of-Cycle-11 Examinations atFarley Unit 2


EPRI Fatigue ManagementHandbook, Vols. 1-4

EPRI sept-95 TR-104534-V1-V4

Computer Manual

EPRI PWR Fuel Cladding Corrosion(PFCC) Model, Vol.2: CorrosionTheory and Rate EquationDevelopment


Evaluation of Cable Polymer AgingThrough Indenter Testing of In-Plantand Laboratory Aged Specimens


Evaluation of EnvironmentalQualification Options and Costs forElectrical Equipment for a LicenseRenewal Period for Calvert CliffsNPP


309

Evaluation of Thermal AgingEmbrittlement for Cast AusteniticStainless Steel Components in LWRReactor Coolnt Systems


Evaluation of Thermal FatigueEffects on Systems Requiring AgingManagement Review for LicenseRenewal for the Calvert CliffsNuclear Power Plant


Generic License Renewal TechnicalIssues Summary


I&C Life Cycle Management PlanMethodology, Vols. 1 and 2

EPRI mars-96 TR-105555-V1-V2

FinalReport

IASCC Susceptibility of Low-FluenceStainless Steels Evaluated by In-Flux Slow Strain Rate Tests


Instrumentation and ControlUpgrade Evaluation Methodology,Vols. 1 and 2


Microstructural Characterization ofRPV Steels: Phase 1 (Joint EPRI-CRIEPI RPV Embrittlement Studies)


Multivariable Analysis of the Effectsof Li, H2 and pH on PWR PrimaryWater Stress Corrosion Cracking


Natural Versus Artificial Aging ofElectrical Components

EPRI sept-97 TR106845 InterimReport

NPP License RenewalEnvironmental Compliance ProgramPlan Manual

EPRI oct-94 TR-104291 FinalReport

NPP License RenewalEnvironmental Life CycleManagement Plan Manual: LicenseRenewal Envrionmental Compliance


Nuclear Plant Life CycleManagement Economics


310

Nuclear Power Plant LicenseRenewal Environmental ComplianceProgram Donald C.Cook NuclearPlant Case Study, Phase 2 -Strategic Planning and DataGathering


Nuclear Power Plant Licenserenewal Environmental ComplianceProgram: Donald C. Cook NPPCase Study. Phase 1


Operating NPP FatigueAssessments


Proceedings: Specialists Meeting onEnvironmental Degradation of Alloy600

EPRI mars-97 TR-104898 Proceedings

Property Damage Risk AssessmentScoping Study for South TexasProject Electric Generating Station


Recommendations for and EffectiveFlow-Accelerated CorrosionProgram

EPRI mars-97 NSAC-202L-R1 FinalReport

Reduction of Oxidation InductionTime Testing to Practice as a LifeAssessment Technique for CableInsulation


RPV Thermal AnnealingAssessment for Two PWR PlantDesigns


Statistical Analysis of SteamGenerator Tube Degradation:Additional Topics


Steam Generator Tube FatigueEvaluation


Stress Corrosion Cracking of RPVSteels


Supplemental RPV SurveillanceProgram Guidelines: Joint EPRI-CRIEPI RPV Embrittlement Studies


311

A Resource Guide to Nuclear PlantLife-Cycle Management

EPRI Report 01/12/1993 TR-103054 Forecasting the useful economic life of a nuclearunit and addressing the complementary issue oflicensing renewal are complex undertakings. Thisguide provides a resource document thatemphasises the technical elements of LCM,focusing on determining adequate maintenanceprograms, and identifying data and recordsnecessary to support them.

Age-Related Degradation InspectionMethod and Demonstration onBehalf of Calvert Cliffs NuclearPower Plant License RenewalApplication

EPRI Report 27/05/1998 TR-107514 Age-related degradation inspections (ARDI) will berequired for license renewal in some cases. Thisreport documents an approach developed by BGEto determine when an ARDI is required andestablishes a standard process to be used todevelop inspection requirements for specificcomponent/ageing mechanisms combinations. TheARDI is then applied to five piping systems and onstructure systems to demonstrate the process.

BWR Containment License RenewalIndustry Report, Rev.1

EPRI Report mars-95 TR-103840 FinalReport

The effect of most age-related degradationmechanisms were found to be covered . Exceptionare: aggressive chemical attack on below-gradeBWR containment concrete structures or corrosionof reinforced steel or embedded steel in below-grade BWR containment steel structures; andcorrosion of inaccessible liner plate regions, or localcorrosion of embedded containment shells andfree-standing steel containments.

BWR Pilot Plant Life ExtensionStudy at the Monticello Plant

EPRI Report 01/05/1987 NP-5181 M This report is one of a series of reportsdocumenting an extensive, 4-year examination ofthe Monticello (BWR) Plant for operation beyondthe initial 40-year licensed term. Ageingmanagement evaluations of critical systems,components, and structures, econmic analyses andanalysis of plant data were undertaken. This studyconfirmed the feasibility of operating Monticello for70 years or more.

BWR Pressure Vessel LicenseRenewal Industry Report, Rev.1


The effect of age-related degradation mechanismswere found to be covered by limits established bythe CLB or managed by current program ofincpection, testing, repair, refurbishment andanalytical evaluation.

312

BWR Primary Coolant PressureBoundary License Renewal IndustryReport, Rev.1



BWR Reactor Pressure VesselInternals License Renewal IndustryReport, Rev.1


The effect of most age-related degradationmechanisms were found to be covered . Theexception were related to creviced locations and todifficulty of inservice examinations for theselocations.

Calvert Cliffs LCM Implementationand Lessons learned

EPRI Report 31/12/1999 TR-107544 BG&E’s Life-Cycle Management Program wasimplemented in 1991, to achieve successful longterm operations of Calvert Cliffs. The reportdocuments the lessons learned by BG&E in thisprocess, both administrative and technical.

Calvert Cliffs License RenewalApplication

EPRI Report 21/11/1997 MI-107543 This is the information required by 10CFR54 inorder to extend the operating license of CalvertCliffs beyond its current 40-year licensed term ofoperation. It is the first complete license renewalapplication of a commercial NPP and likely to bethe first license renewal application submitted to theNRC.

Calvert Cliffs License RenewalApplication Technical Basis (Rev.1)

EPRI Report 31/12/1999 TN-110163 CD-Romlimited toEPRINPGmember

This product is a user’s guide for a CD-ROM whichcontains 100 documents consisting of the CalvertCliffs License Renewal Methodology and NRCSER, procedures for evaluation of systems andstructures, Ageing management Review Reports forsystems, structures and commodities, licenserenewal application technical reports...

Calvert Cliffs License RenewalApplication: NRC Final SafetyEvaluation

EPRI Report 30/06/2001 TR-107542 Issuance of the NRC Final Safety Evaluation for theCalvert Cliffs license renewal application maks asignificant milestone in the demonstration of aviable and predictanle process for nuclear powerplant license renewal and preserving the nuclearoption in the United States.

Calvert Cliffs NPP Integrated PlantAssessment Methodology

EPRI Report juil-97 TR-104734 FinalReport

This report documents the plant-specific processused for conducting the integrated plantassessment for aging and time-limited ageinganalyses. This report also incorporates theresolution of comments and request for additionalinformation for the NRC final safety evaluation.

313

Calvert Cliffs NPP life cyclemanagement /license renewalprogram: containment systemcomponent evaluation and programevaluation


Calvert Cliffs NPP Life CycleManagement-License RenewalProgram Structure Scoping andAging Management Review

EPRI Report mars-96 TR-105420-V1,V2,V3

FinalReport

This report documents the identification ofstructures and their components within the scope oflicense renewal for Calvert Cliffs and the results ofthe aging management evaluation for thesestructures and components. There are threevolumes which provide the results of scoping andcomponent aging management evaluations of sixstructures: containment, condensate storage tank,fuel oil storage tank enclosure, turbine building,intake structure, and auxiliary building structure.

Class 1 Structure License RenewalIndustry Report, Rev.1


The effect of most age-related degradationmechanisms were found to be covered . Exceptionare: aggressive chemical attack on below-gradeClass I concrete structures, corrosion ofinaccessible or below grade, Class I steelstructures, and intergranular stress corrosioncracking or crevice corrosion of austenitic stainlesssteel liners of Class I concrete structures ouaustenitic stainless steel tanks.

Generic License Renewal TechnicalIssue Summary (Rev.1)

EPRI Report 01/09/1999 TR-110902-SI This report documents industry technical positionson generic license renewal technical issues. Foreach issue, the background is reviewed, theindustry position is presented, and the technicalbasis for the industry position is provided. Thisreport updates a previous report (EPRI TR-107521).

LCM primer EPRI Report 01/12/1998 TR-106109 The LCM Primer provides information on thebackground, developments, issues, and sources ofinformation on Life Cycle Management. LCM is thesystematic integration of activities that determinethe useful life. It combines a diverse set of issues:economics, operations, material aging, licensing,environmental, public relation, spent fuel disposal,low level waste and others.

314

License renewal Industry ReportSummary

EPRI Report 22/06/1998 TR-104305 This report summarises the conclusions andtechnical findings of the 10 license renewal industryreports, and the NRC staff position on these issues.It includes a comparison with the results reported inNureg 1557, and identifies significant differences onlicense renewal technical issues.

License Renewal TechnicalImplementation Guide

EPRI Report 01/08/1995 TR-105090 This report documents the initial results of the NEILicense Renewal Implementation Guideline TaskForce over the period August 1994 to July 1995 todevelop guidance for complying with the technicalrequirements of the License Renewal Rule 10 CFR54. This EPRI report also contains an "Identificationof potential time-limited ageing analyses inherent inthe common codes and standards and genericregulatory compliance documents for NPPs".

Low-voltage environ-mentallyqualified cable license renewalindustry report, Rev.1


The effect of age-related degradation mechanismswere found to be covered by limits established bythe CLB or managed by current program ofinspection, testing, repair, refurbishment andanalytical evaluation.

LWR Plant Life Extension EPRI Report 01/01/1987 (592p) This report documents interim results of two lifeextension pilot studies at Surry and Monticello.

Monticello Lead Plant LicenseRenewal Project: Summary Report

EPRI Report janv-97 TR-103963 FinalReport

This report summarises license renewal evaluationactivities from the Monticello Lead Plant Licenserenewal project that were conducted betweenOctober 1988 and December 1992,The lessonslearnd during this first plant specific license renewalevaluation were instrumental in leading to a changein the License Renewal Rule (10CFR54).

Oconee License RenewalApplication

EPRI Report 01/03/2001 TR-111030-CD This brief report and CD-ROM contain the completeOconee License Renewal Application submitted tothe NRC in early July, 1998, It is the secondLicense Renewal Application submitted for acommercial nuclear power plant.

315

PWR Containment StructuresLicense Renewal Industry Report,Rev.1


The effect of most age-related degradationmechanisms were found to be covered . Exceptionare: aggressive chemical attack on below-gradePWR containment concrete structures or corrosionof reinforced steel or embedded steel in below-grade PWR containment steel structures; andcorrosion of inaccessible, or below-grade steel,steel structures including both liners for PWRconcrete containment and freestanding steel shellsfor PWR steel containment.

PWR pilot plant life extension studyat Surry Unit 1: Phase 2

EPRI Report 01/03/1989 NP-6232-SD(451p)

PWR Reactor Coolant SystemLicense Renewal Industry Report,Rev.1



PWR Reactor Pressure VesselInternals License Renewal IndustryReport, Rev.1


The effect of most age-related degradationmechanisms were found to be covered . Theexception were related to stress relaxation effectson bolts, pins and fasteners that are not readilyaccessible for visual examination, and IASCC ofaustenitic stainless steel internals subjected torelatively-high levels of tensile stress, neutronfluence and aggressive coolant environment.

PWR Reactor Pressure VesselLicense Renewal Industry Report,Rev.1



Review of NRC GenericCommunications for LicenseRenewal

EPRI Report 12/03/1998 GC-107944 This report documents a review of all NRC genericcommunications to determine those that involveboth long-lived passive components and structuresrequiring ageing management review and an age-related degradation mechanism or ageing effect,and therefore may be potentially applicable tolicense renewal.

316

Risk-Informed Inservice InspectionEvaluation Procedure

EPRI Report May 1997June 1996Mai 1997

TR-106706 Interimreport

This report describes evaluation procedures forusing risk to define inspection locations for safetysignificant piping in nuclear power plants. PSAinsights, deterministic evaluations, and plantinservice experience are integrated in a practicalformat. Procedures in the report are intended tosupport ongoing pilot plant application studies.Lessons learned will be incorporated into a finalreport.

Surry Unit 1 Plant Life ExtensionProgram

EPRI Report 01/09/1987 NP-5289SP This is one of several reports documenting theresults of a life extension study for Virginia Power’sSurry Unit 1. These studies determined that a 60-year operating period was technically achievableand economically attractive.

Time Limited Ageing Analysis(TLAA)

EPRI Report 01/12/1999 TR-110042 This report documents the identification andevaluation of time ageing analysis for Plant Hatch.A complete design basis review will be performed inorder to identify calculations, evaluations, analyses,committments and any other documentation, whichinvolves the assessment of the functionality of aplant system, structure or component for a specifiedtime period.

Utility Activities for Nuclear PowerPlant Life Cycle Management andLicense Renewal

EPRI Report sept-95 TR-104751 FinalReport

This is a key EPRI LCM report which providesguidance to nuclear utilities on steps to take,industry activities undertaken and in progress, andproducts developed for LCM and License Renewalactivities. It provides information to assist utilities inestablishing LCM programs.

Research and Training programmein the field of nuclear energy (1998to 2002)

EURATOM

Answer to the National ProfilQuestionnaire on Nuclear PlantAgeing, Life Management, and PlantLife extension -Belgium


01/04/1992

AMAT Guidelines IAEA March 1999 IAEA ServicesSeries N°4

Assessment and management ofageing of major nuclear power plantcomponents important to safety:PWR vessel internals.

IAEA Report October 1999 IAEA TECDOC -1119

317

Assessment and management ofageing of major nuclear power plantcomponents important to safety:Steam Generators

IAEA Report November1997

TECDOC-981

Assessment and Management ofAgeing of Major Nuclear PowerPlants Components Important toSafety: Concrete ContainmentBuildings

IAEA Report June 1998 TECDOC-1025

Assessment and Management ofAgeing of Major Nuclear PowerPlants components important toSafety: Pressurized Water ReactorPressure Vessels

IAEA Report October 1999 IAEA - Tecdoc1120

Code on the Safety of NuclearPower Plants: Operation

IAEA Report 1988 Safety Series No50-C-0 (Rev.1)

Co-ordinated Research Programme(CRP) on Management of Ageing ofIn-Containment Instrumentation andControl Cables

IAEA Report

Data collection and Record keepingfor the management of NuclearPower Plant Ageing - A SafetyPractice

IAEA Report December1991

Safety Series No50-P.3

Equipment Qualification inOperational Nuclear Power Plants

IAEA Report 1998 Safety ReportsSeries No3

Evaluation of the safety of operatingnuclear power plant built to earlierstandards - A common basis forjudgment

IAEA Report Safety ReportsSeries No 12

Implementation and Review of aNuclear Power Plant AgeingManagement Programme

IAEA Report avr-99 Safety Series No15

International database on ageingmanagement and life extension -Database specification

IAEA Report 1994 IWG-LMNPP-94/6

Management of Nuclear PowerPlants for Safe Operation: A safetyGuide

IAEA Report 1994 Safety Series No50-SG-09

318

Methodology for the management ofageing of nuclear power plantcomponents important to safety

IAEA Report juil-92 TechnicalReports SeriesNo 338

Periodic Safety of OperationalNuclear Power Plants: A safetyGuide

IAEA Report 1984 Safety Series No50-SG-12

Pilot Studies on Management ofAgeing of Instrumentation andControl Cables

IAEA Report TECDOC-932

Pilot studies on Management ofAgeing of Nuclear Power PlantComponents - Results of Phase I

IAEA Report October 1992 TECDOC-670

Regular meeting of the int. WG onLife management of NPP (IWG-LMNPP)

IAEA Report 25-28 May,1999

Minutes ofmeeting

Review of Selected Cost Drivers forDecisions on Continued Operation ofOlder Nuclear Reactors - SafetyUpgrades, Lifetime Extension,Decommissioning

IAEA Report May 1999 TECDOC-1084

Safe Management of the OperatingLifetimes of Nuclear Power Plants

IAEA Report November1999

INSAG 14

Safety Aspects of Nuclear PowerPlant Ageing

IAEA Report 1990 TECDOC-540

Operating experience to apply toadvanced light water reactor designs

INPO Report March 1996 INPO 93-004,Rev. 04

Limiteddistribu-tion

The information in this document complements theUtility Requirements Document and other ongoingeffort by designers to review and apply valuablelessons learned from industry operating experienceto advanced LWR designs. Designers shouldreview the problems addressed in the reports andstrive to overcome them in the designs.

AMES REFERENCESLABORATORY JRC-IAM/ECNPETTEN

JRC-IAM/ECN juin-96 CD-NA-16409-EN-C

A COMPARASION OF WESTERNAND EASTERN NUCLEARREACTOR PRESSURE VESSELSTEELS

L.M.Davies avr-97 CD-NA-17327-EN-C

SURVEY OF EXISTING, PLANNEDAND REQUIERED STANDARDS

MPA STUGART déc-95 CD-NA-16313-EN-C

319

Boraflex Degradation In Spent FuelPool Storage Racks

NCR juin-96 Gl 96-04

Degradation of Control Rod DriveMechanism And Other VesselClosure Head Penetrations

NCR avr-97 Gl 97-01

Degradation Of Scram Solenoid PilotValve Pressure And ExhaustDiaphragms

NCR oct-94 IN 94-71

Electrical Penetration AssemblyDegradation

NCR avr-93 IN 93-025

Potential For Degradation Of TheEmergency Core Cooling SystemAnd The Containment Spray SystemAfter A Loss Of Coolant AccidentBecause Of Construction AndProtective Coating Deficiencies AndForeign Materials In Containment

NCR juil-98 Gl 98-04

Snubber Lubricant Degradation InHigh-Temperature Environments

NCR juin-94 IN 94-48

Unexpected Degradation Of LeadStorage Batteries

NCR avr-95 IN 95-21

Industry Guideline for Implementingthe Requirements of 10CFR Part 54- The License Renewal Rule

NEI mars-96 NEI 95-10

Industry Guidelines for Implementingthe Requirements of 10CFR Part 54- The license Renewal Rule

NEI Report Rev.1January 2000

NEI 95-10 Approach to implement the requirements of 10CFRpart 54, the license renewal rule, based on industryexperience and expertise in implementing the LRR.

Application of NUREG/CR-5999interim fatigue curves to selectedNuclear Power Plant components

NRC mars-95 NUREG/CR-6260

Assessment of thermalembrittlement of Cast StainlessSteels

NRC déc-93 NUREG/CR-6177

Degradation in small-radius U-bendregions of steam generator tubes

NRC juil-98 IN-97-26

Degradation of ventilation systemcharcoal resulting from chemicalcleaning of steam generators

NRC sept-95 IN 95-41

Detection of pump degradation NRC août-95 NUREG/CR-6089Draft Regulatory Guide DG-1047 NRC August 1996 DG-1047

320

Interim fatigue design curves forcarbon, low-alloy and austeniticstainless steels in LWRenvironments

NRC avr-93 NUREG/CR-5999

Working Draft Standard Review Planfor License Renewal

NRC April 2000 WD-SRP-LR

1995 Consensus Document onSafety of European LWR


1997 EUR 16803 EN

Common position of Europeanregulators on qualification of NDTsystems for pre- and in-serviceinspection of Light Water reactorcomponents


1997 EUR 16802 EN

Periodic safety reviews of NPPs inEC Member states, Finland, Swedenand Switzerland: a review of currentpractices


1995 EUR 13056 EN

Seismic re-evaluation of operatingnuclear power plants in Europeancountries - Comparative study ofnational practices


1996 EUR 16245 EN

BWR Containments LicenseRenewal Industry Report

NUMARC juil-94 IR-90-10

BWR Pressure Vessel LicenseRenewal Industry Report


BWR Primary Coolant PressureBoundary License Renewal IndustryReport


BWR Vessel Internals LicenseRenewal Industry Report


Class 1 Structures License RenewalIndustry Report


Low Voltage Environmentally-Qualified Cable License RenewalIndustry Report


PWR Containment StructuresLicense Renewal Industry Report


PWR Pressure Vessel InternalsLicense Renewal Industry Report


321

PWR Reactor Coolant SystemLicense Renewal Industry Report


PWR Vessel License RenewalIndustry Report


License Renewal Industry Reports Numarc Report 1990 IR-90-01 to 10 The ten Industry reports (IR) provide a generictechnical basis for evaluation of Nuclear Powerplants components for license renewal

Refurbishment costs of nuclearpower plants

OECD Report January 199976p

NEA/NDC/DOC(99)1

Restric-ted

It is of utmost importance for electric utilities toprove the competitiveness of nuclear power. Thestudy performed is a component of OECD/NEAprogramme on Plant Life Management and theobjective was to collect and evaluate nuclear powerplant refurbishment cost data and experience. Thisinformation is useful to reactor operators faced withnuclear plant life cycle evaluation.

Regulatory Aspects of ageingReactors

OECD/CNRA March 1999 NEA/CNRA/R(99)1

Report on future nuclear regulatorychallenges

OECD/CNRA November1997, 47p

Finaldraft

For nuclear regulatory bodies, the first challenge isto ensure that economic pressures do not erodenuclear safety. The report identifies the challengesthat regulators are likely to be confronted with overthe next ten years

Development priorities for NDE ofconcrete structures in nuclear plants(NEA Workshop)

OECD/CSNI Risley, UK,Nov. 19971998

NEA/CSNI/R(97)28

FALSIRE: phase 1: CSNI project forfracture analyses of large-scaleinternational reference experiments -Comparison report

OECD/CSNI 1997 OECD/GD(97)24

FALSIRE: phase 2: CSNI project forfracture analyses of large-scaleinternational reference experiments

OECD/CSNI 1996 NEA/CSNI/R(96)1OCDE/GD(96)187

Fatigue crack growth benchmark OECD/CSNI 1997 NEA/CSNI/R(97)8

International Workshop on Aged andDecommissioned Material Collectionand Testing for Structural IntegrityPurposes

OECD/CSNI 1995; Mol,Belgium1996

NEA/CSNI/R(95)17OCDE/GD(96)10

322

Joint WANO/OECD-NEA Workshop:Prestress loss in NPP containments

OECD/CSNI 1997:Poitiers,France1997


Leak before break in reactor pipingand vessels: specialists meeting

OECD/CSNI 1995: Lyon,France1996, Vol 1-3


NDE technique capabilitydemonstration and inspectionqualification: proceedings of the jointEC OCDE IAEA specialists meeting

OECD/CSNI 1997: Petten,TheNetherlands1997

NEA/CSNI/R(97)1EUR 17354 EN

Probabilistic structure integrityanalysis and its relationship todeterministic analysis

OECD/CSNI 1996:Stockholm,Sweden1996


Report of the task group on theseismic behaviour of structures:status report


Report of the task group reviewingnational and international activities inthe area of ageing of nuclear powerplant concrete structures


Report on round robin activities onthe calculation of crack openingbehaviour and leak rates for smallbore piping components


Seismic shear wall ISP: NUPEC’sseismic ultimate dynamic responsetest: comparative report


State of the art on key fracturemechanics aspects of integrityassessment


Workshop on Reactor CoolantSystem Leakage and FailureProbabilities

OECD/CSNI 1992; Köl,Germany1995


Status report on Nuclear PowerPlant Life management

OECD/NEA 10 Feb - 2000 ThirdDraft,ver.4

323

Aging Management Guideline forCommercial Nuclear Power Plants -Power and Distribution Transformer

SANDIALaboratories

1993 SAND93-7068

Aging management guideline forcommercial Nuclear Power plants -tanks and pools

SANDIALaboratories

févr-96 SAND96-0343 /UC-523

CONVERSION TABLE OFMATERIAL DAMAGE INDEXATIONFOR ALL DIFFERENT EUROPEANREACTOR TYPES

TECNATOM déc-98 CD.NA-18693-EN-C

DOSIMETRY AND IRRADATIONPROGRAMMES OF AMESEUROPEN NETWORK

TECNATOM déc-97 CD.NA-17744-EN-C

DOSIMETRY AND NEUTRONTRANSPORT METHODS FOR RPV

TECNATOM nov-96 CD-NA-16470-EN-C

SURVEY OF NATIONALREGULATORY REQUIREMENTS

TRACTEBEL juin-95 CD-NA-16305-EN-C

Comportamiento frente a lairradiacion neutronica de losmateriales de vasijas de reactoresde agua ligera

UNESA 1995 ISBN: 84-7834-186-2

Spanish The purpose of this document is to present theactivities, results and conclusions of the ResearchProject Behaviour with regard to Neutron Irradiationof Light-Water Reactor Vessel Materials. Itsobjective is the optimisation and enhancement ofthe embrittlement surveillance procedures, directlyrelated to an adequate management of theremaining life of the reactor.

Analisis fenomenos degradatoriosen generadores de vapor

UNESA /CIEMAT February1993

GVR-CI-IT-03202 Spanish The document covers a survey on the degradationmechanisms which can affect the steam generatorsat Vandellos II NPP. It covers several problems, thereasons for formation and propagation,identification and follow-up parameters, and controland mitigation methods.

Análisis Fenómenos Degradatoriosen Generadores de Vapor

UNESA/CIEMAT FEB-93 GVR-CI-IT-03202 FINAL

Análisis Fenómenos Degradatoriostuberías Recirculación ReactoresBWR

UNESA/CIEMAT NOV-92 GVR-CI-IT-03201 FINAL

Estudio Fenómenos Degradatorios.Dossier Bombas, Turbinas yCompresores

UNESA/EA JUN-93 GVR-EA-IT-03312

FINAL

324

Estudio Fenómenos Degradatorios.Dossier Componentes de ventilación


FINAL

Estudio Fenómenos Degradatorios.Dossier Contenciones Metálicas

UNESA/EA APR-93 GVR-EA-IT-03308

FINAL

Estudio Fenómenos Degradatorios.Dossier Equipo Eléctrico


FINAL

Estudio Fenómenos Degradatorios.Dossier Generador Diesel


FINAL

Estudio Fenómenos Degradatorios.Dossier Instrumentación y Control


FINAL

Estudio Fenómenos Degradatorios.Dossier internos Vasija BWR

UNESA/EA MAY-93 GVR-EA-IT-03306

FINAL

Estudio Fenómenos Degradatorios.Dossier Tuberías y Válvulas


FINAL

Estudio Fenómenos Degradatorios.Dossier Vasija Reactor BWR


FINAL

Evaluación de Mantenimiento deCNSMG Vol. I: Mecánico

UNESA/EA MAY-94 GVR-EA-IT-04402-1

FINAL

Evaluación de Mantenimiento deCNSMG Vol. II: Eléctrico


FINAL

Evaluación de Mantenimiento deCNSMG Vol. III: I&C


FINAL

Evaluación de Mantenimiento deCNV II Vol. I: Mecánico


FINAL

Evaluación de Mantenimiento deCNV II Vol. II: Eléctrico


FINAL

Evaluación de Mantenimiento deCNV II Vol. III: I&C


FINAL

Guia de evaluacion de practicas demaintenimiento

UNESA/EA September1993

GVR-EA-GU-04101 Rev.1

Spanish Demonstrating that ageing and performance arebeing effectively managed is a key element of asuccessful lifetime management program. Thepurpose of the evaluation guide is to provide astandardised approach for the maintenanceeffectiveness evaluations. Three parts aredescribed: identification of degradationmechanisms, collection of information, maintenanceevaluation.

325

Guia de seleccion de componentesimportantes para la gestion de vidade centrales nucleares

UNESA/EA January 1993 GVR-EA-GU-00101 Rev.1

Spanish The objectives of this document is to provide astandard method for selecting and prioritising plantstructures, systems and components important toevaluating the lifetime of LWR nuclear plantcomponents. These components have the greatestsensitivity to ageing, and O&M costs will dictate thefeasibility of Life Cycle Program for the plant.

Hojas de Datos de Prácticas deMantenimiento de CNSMG Vol II.Eléctrico

UNESA/EA MAR-94 GVR-EA-IT-04302-2

FINAL

Hojas de Datos de Prácticas deMantenimiento de CNSMG Vol. III.I&C


FINAL

Hojas de Datos de Prácticas deMantenimiento de CNSMG Vol.IMecánico


FINAL

Hojas de Degradación deComponentes de CNSMG Vol. I.Mecánico


FINAL

Hojas de Degradación deComponentes de CNSMG Vol. II.Eléctrico


FINAL

Hojas de Degradación deComponentes de CNSMG Vol. III.I&C


FINAL

Recomendaciones de vigilancia ymonitorización. Definición dealcances


FINAL

Requisitos fundamentales devigilancia y monitorización


FINAL

Selección de Componentes CríticosC.N.S.M.G.

UNESA/EA JAN-93 GVR-EA-IT-02202

FINAL

Selección de Componentes CríticosC.N.V.II

UNESA/EA JAN-93 GVR-EA-IT-02102

FINAL

Cuantificación Modos Degradaciónpor irradiación, envejecimientoTérmico y fatiga

UNESA/FRAMATOME

FEB-93 GVR-FI-IT-03206 FINAL

Fenómenos Degradatorios.Componentes Pesados

UNESA/FRAMATOME


326

Fenómenos Degradatorios.Generadores de Vapor yPresionador

UNESA/FRAMATOME

JAN-93 GVR-FI-IT-03203 FINAL

Fichas de Evaluación deFenómenos Degradatorios

UNESA/FRAMATOME


Informe de Síntesis para laEvaluación de Vida Remanente delos circuitos

UNESA/FRAMATOME

MAR-93 GVR-FI-IT-03205 FINAL

Estudio Fenómenos Degradatorios.Dossier Cambiadores de Calor yDepósitos

UNESA/INITEC MAY-93 GVR-IN-IT-03314 FINAL

Estudio Fenómenos Degradatorios.Dossier Estructuras Hormigón

UNESA/INITEC APR-93 GVR-IN-IT-03307 FINAL

Estudio Fenómenos Degradatorios.Dossier Generador de Vapor


Estudio Fenómenos Degradatorios.Dossier Internos Vasija PWR


Estudio Fenómenos Degradatorios.Dossier Presionador


Estudio Fenómenos Degradatorios.Dossier Soportes Principales


Estudio Fenómenos Degradatorios.Dossier Turbogrupo

UNESA/INITEC JUN-93 GVR-IN-IT-03311 FINAL

Estudio Fenómenos Degradatorios.Dossier Vasija Reactor PWR


Hojas de Datos de Prácticas deMantenimiento de CNV II


Hojas de Degradación deComponentes de CNV II


Application of SAT: systematicapproach to training of nuclearpower plant operators

UNIPEDEPublications

1997 - 17p -EN

01004Ren9757 Public The group "Training of operational personnel innuclear power station composed of experts andmanagers responsible for operator training anddevelopment from 12 european companies, hasexamined the objectives, rules, methods, policiesand approaches to training activities. This hasresulted in a series of documents covering aspectsessential for promoting safety and improvingperformance of the nuclear generating park.

327

At your own risk - an inquiry aboutrisk management in the ESI

UNIPEDEPublications

1996 - 40p -EN

00300Ren9609 Public The paper explains what is understood by RiskManagement and gives an overview of the resultsof the inquiry carried out in 1995 on how RiskManagement is being applied in the ElectricityIndustry. The analysis was based on 85 replies to aquestionnaire, received from 81 electricitycompanies in 31 countries.

Condition monitoring for powerplants

UNIPEDEPublications

1997 - 20p -EN

01002Ren9751 Public Systems for condition monitoring of equipment andcomponents in nuclear and non-nuclear generatingstations are reviwed. The report covers monitoringof vibrations (rotating components, structures),loose parts, leakage, fatigue, noise; electrical andI&C systems are also discussed. The reportincludes the resulys of a survey among utilities onthe present status of the art and addresses futuretrends.

Experience and progress in nuclearpower production of Unipedemember countries

UNIPEDEPublications

1997 - 14p -EN

01000Ren9793 Public At the Unipede congress (Montreux, May 1997), theUnipede nuclear generation and thermal generationstudy committees held a joint session on "Poweradvances in their operations". This sessiondocument covers power advances and trends inimproved availability, power upgrading, safetyperformance, plant life extension, modernisationprogrammes, radioactive operational waste, spentfuel and decommissioning.

Primary circuit chemistry of WesternPWR and VVER plants

UNIPEDEPublications

1996 - 148p -EN

02004Ren9653 Public The report contains detailed informations on thedesign and operating experiences of both PWRsand VVERs. It is a simultaneous updating of thestate of the art for PWR/VVER primary circuitchemistry and a document that can be used as abasis for defining the chemical requirements for thedesign of primary circuit of future PWRs. It isUnipede’s aims that training should be provided toimprove the safety culture in European powerreactors. This report will assist in this aim.

328

Safety evaluation of nuclear powerplants designed to earlier standards

UNIPEDEPublications

1996 - 22p -EN

01005Ren9636 Public Nuclear power plants designed to earlier standardsmust have an adequate level of safety. Mostregulators are now seeking additional assurancethat NPPs sre continuing to be operated atreasonable levels of safety when compared with thesafety standards which would apply to a NPPdesigned and constructed today. This is reflectedby principle 25 of the Safety Fundamentals of theIAEA.

The European Utility Requirement(EUR) document in 1997: progressand near term objectives"

UNIPEDEPublications

1997 - 4p -EN

01000Ren9797 Public The objective of the report is to update previousversions of defining common ground, acceptable toutilities, the public and the administrations, on thedesign of standard LWR nuclear power plants inWestern Europe. The safety approaches, targetsand criteria of the future plants, their designconditions, their performance targets, their systemsand equipment specifications are being harmonisedunder the leadership of the electricity producers.

COMPARISON OF SCIENTIFICBASIS OF RUSSIAN ANDEUROPEAN APPROACHES FOREVALUATING IRRADATIONEFFECTS IN REACTORPRESSURE VESSELS

VTT févr-95 CD-NA-16279-EN-C

IRRADATION EMBRITTLEMENTMITIGATION

VTT sept-94 CD-NA-16072-EN-C

STATE OF THE ART REVIEW ONTHERMAL ANNEALING

VTT déc-94 CD-NA-16278-EN-C

Aging and service wear of air-operated valves used insafety-related systems at nuclear power plants

août-94 NUREG CR-6016

Aging and service wear of spring-loaded pressure reliefvalves used in safety-related systems at nuclear powerplants

mars-95 NUREG CR-6192; ORNL-6791

Aging assessments of bistables andswitches in nuclear power plant

janv-93 NUREG CR-5844/ BNL-NUREG52318

329

Aging assessments of lange electricmotors in nuclear power plant

mars-96 NUREG CR-6336/ BNL-NUREG-52460

Aging of safety class 1etransformers in safety systems ofnuclear power plants

nov-97 NUREG CR-5753

Component unavailability versus inservice test (IST)interval: evaluations of component aging effects withapplications to check valves

juil-97 NUREG CR-6508; ORNL- 6909

Development priorities for NDE ofconcrete structures in nuclear plants

1998 NEA/CSNI/R(98)6

Draft conclusions and recommendations from theworkshop on the finite elements analysis of degradedconcrete structure

April 1999

Draft conclusions andrecommendations on containmenttendon prestress loss

April 1999

Effect of dynamic strain aging ofnuclear ferritic piping at LWRtemperatures

juil-94 NUREG CR-6226; BMI-2176

Estimation of FractureThougness of Cast Stainless SteelsDuring Thermal Aging in LWR Systems

août-94 NUREG CR-4513

Evidence aging effects on certainsafety-related components

janv-96 NUREG CR-6442 ; INEL-95/0654 ;NEA/CSNI/R(95)9

Evidence of Ageing effects on certain safety relatedcomponentsVolume 1: summary and analysisVolume 2: contributions

Septembre1995Vol.1: 68p

NEA/CSNI/R(95)9

Experience with thermal fatigue in LWR piping caused bymixing and stratification

December1998

NEA/CSNI/R(98)8

Long-term aging and loss-of-coolantaccident (LOCA) testing of electricalcables

oct-96 NUREG CR-6202 ; SAND-0485 ; IPSN 94-03

330

Nuclear power plant gennericlessons learned (GALL)

déc-96 NUREG CR-6490; ANL 96/13 V 1

Operation and MaintenanceExperience with computer-basedSystems in NPP's

Septembre199853p

NEA/CSNI/R(97)23

PISC III: Final report 1998 NEA/CSNI/R(98)9

Finalreport

Plant Ageing Management - Providing a technical basisfor long-term operation of light water reactors

May 1999 CSNI-PWG3TechnicalPositionDocument

Draft

PLIM Workshop - 6th Meeting of the Expert Group onNuclear Power Plant Life Management

June 1997229p

NEA/SEN/NDC(97)11, Rev.1

Relation of ageing and seismicengineering

June 1999 OECD/PWG3 Draft

Safety practices. Data collection and record keeping forthe management of NPP ageing

janv-91 Ref. EDF:E1992E150084

Status report on seismic-re-evaluation

NEA/CSNI/R(98)5

Survey of organic components innuclear power plants

1998 NEA/CSNI/R(98)7

331

Appendix_6: Publications

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ANERI Current Status of Nuclear PowerPlants in Japan (copy of slides)

1 October1997

IAEA IAEA Program on the SafetyAspects of Nuclear Power PlantAgeing

1st meeting ofExpert Group onPlant LifeManagementMeeting, Paris, 18-19 June 1991

17 June1991

JAPEIC Nuclear Power Plant LifeManagement in Japan (Copy ofslides)


17 June1991

OECD/NEA

Brief Overview of the Plant LifeManagement decision makingprocess


EPRI Joint DOE-EPRI strategicresearch and development planto optimise US NPP - Volume 1-Executive summary and Volume2 - table of content.

EPRI 20 March1998

EDF Life time management/Lifeextension (copy of slides)

European ExecutiveAdvisory Board, N°5- Panel Discussionon the Topic "LifeExtension to 60years", 2/10/97

2 October1997

NOK Aging Management KKB (copy ofslides)


2 October1997

332

IAEA Implementation and review ofnuclear power plant ageingmanagement program - A safetyreport (Final Draft)

IAEA 1 March1998

IAEA IAEA Consultants’ Report on theMeeting on Nuclear Power PlantAgeing and Life Management

IAEA, Vienna,Austria, April 17-21,1989

17 April1989

French 900 MW PWR unitIncident on Seismic Stops (Copyof slides)

Meeting on theseismic behaviour ofstructures,December 2-3,1996,OECD/NEA/CNRA/PWG-3

2December

1996

ASME Draft Non-Mandatory Appendix xEnvironmental effects oncomponents

Members SubGroupDesign (SC III)

5 May1997

RSWG Technical Meeting on AgeingManagement (copy of slides)

RSWG Task Forceon Ageing, Brussels

24 March1998

Stokoe T. MagnoxElectric

Magnox Electric Approach tocontinued operation of nuclearpower plants (OECD/NEA -International Workshop OnNuclear Power Plant LifeManagement)

6th Meeting of theExpert Group onNuclear Power PlantLife Management,Paris, April 14-15,1997

14 April1997

Aguado M. Seguimiento de vida remanenteen base a sistemas inteligentesde gestión de inspecciones

21 Sociedad NuclearEspañola AnnualMeeting

25-28October95

Tarragona (Spain)

Aguado M. Inspección y analisis deintegridad. Herramientas básicaspara la gestión de vidaremanente

International Seminar"Alargamiento de Vida deEquipos e Instalaciones"

February1995

Bilbao(Spain)

Aguado M. Ageing, Surveillance andRemanent Lifetime EvaluationBased on an Intelligent InspectionPlanner and ManagementSystem

PLIM +PLEX 95 Page 489,PosterSession 1

27-30November95

Nice(France)

333

Aguado M. , I.Marcelles

Towards the integration ofremanent lifetime andmaintenance based on conditionmonitoring techniques

IAEA SpecilistMeeting

Page 121 2-5 June1998

Lyon(France)

Aguado M., C. Cueto Seguimiento de la degradación yevaluación de vida remanente decomponentes de centralesnucleares

Revista de laSociedad NuclearEspañola

February1997

Madrid

Aguado M., I.Marcelles

Hacia la Integración deHerramientas en los Ambitos deGestión de Vida, Mantenimientoe Inspección


Page 455,Session23-10

5-7November97

LaCoruña(Spain)

Aguado M., I.Marcelles

Towards the integration ofremanent lifetime, maintenanceand inspection tools

Top-Safe 15-17 April1998

Valencia(Spain)

Aguado M., J. Ortega,I. Marcelles

Improvement of NPP Lifemanagement using advancedNDT techniques: A look to thefuture

PLIM +PLEX 97 Page 231,Session II:Monitoring,Surveillance,Inspection

8-10December97

Prague (Czech Republic)

Ahmed I., W.M. Butt,Z.H. Siddiqui, J. Iqleem

Kanupp Plant life extension at Kanupp: anupdate

Nuclear EngineeringInternational

1 June1997

Allen R.P., J.J. Burns Shippingport station agingmanagement lessons

Aging research inf.conf.

Nureg/CP-0122-Vol.2 p.151-164

24-27 Mar1992

Rockville

Alt M., M. Fuchs, H.Krapf, U. Peter, H.Schalk, M. Seevers, M.Wenk

GKN, PE,KKG,RWE-E,KKP, HEW,KWO

Plant Life Management inGerman Nuclear Power Plants

VGB TechnicalAssociation of LargePower PlantOperators

1Septembe

r 1997

Anon NEI (UK) Dukovany: investing in a longterm future for V-213s

NEINBFISSN: 0029-5507

p. 20-22 Dec. 1994

Antonov A.V., A.V.Dagaev, I.S. Volnikov

INPE The development of techniquesfor determining the residual lifetime prediction of NPP equipment

7th InternationalConference OnNuclear Engineering(ICONE)

Paper7119

April 19-23, 1999

Tokyo,Japan

334

Aubry P., J-P. Goffin GecAlsthom,TEE

Rénovation des rotors B.P. desturbines nucléaires - Application àdeux turbines à vapeur de 1000MW en Belgique

Conférence SFEN:Le parc Nucléaire:sa gestion dans ladurée, Paris, 10-11décembre 1996

10December

1996

Aye L. EDF Primary System maintenanceStrategy

ABB Nuclear ServiceSymposium,Saltsjöbaden, June16-18

Badlan M. , R.González, L. Luccardi

Eliminación de TensionesResiduales por ProcedimientosMecánicos (MSIP). Un Mediopara Prevenir la Corrosión Bajotensión en las Cabezas de VasijaTipo PWR



5-7November97

LaCoruña(Spain)

Ballesteros A., L.Debarberis, W.Voorbraak, C. Sciolla,M. Valo, T. Lewis, H.Aït Abderrahim

TecnatomJRC-IAM,NRG, VTT,BNFL,SKC-CEN

RPV Dosimetry Activities: ProjectAMES Dosimetry and MADAM

FISA 99 - EUResearch in ReactorSafety

29 Nov,-1Dec, 1999

Luxembourg

Bartonicek J., O.Jonas, F. Schoeckle

GKKNeckarJonas inc.AMTEC

Quantification of the safety statusof PWR steam generators

ASME Conf. PVP 332,p. 59-65

21-26 Jul1996

Montreal

Bartonicek J., W. Zaiss NeckarWestheim

Guarantee of component integrityas a basis of aging power plantlife management

KT/KTA Winterseminar: Agingmanagement inNPPs

pp. 25-65(German)

25-26 Jan1996

SalzgitterGermany

Bedzikian G., Ensel C.,Churier-Bossennec H.

EDFFramatome

Prise en compte de la cinétiquedu vieillissement thermique dansles études de durée de vie

RGN, RevueGénérale Nucléaire

nov-99 France

Bedzikian G., J.P.Massoud, S. Jayet-Gendrot, H. Churier-Bossennec, P. LeDelliou

EDF Life evaluation of cast duplexstainless steel elbows in FrenchPWRs


Paper7320

April 19-23, 1999

Tokyo,Japan

Berto D.S. ABB-CE Demonstrating safety duringlicense renewal should not be alarge task

1993 PVP Conf. p. 33-39 25-29 Jul1993

New-York

Bieniussa K.W., H.Reck

GRS Evaluation of piping damage inGerman nuclear power plants

Nucl. Eng. AndDesign, 171

pp. 15-32 1997

335

Bollini G.J., J. Navarro Optimización de Técnicas deInspección en Servicio comoPaso Previo para su Validación



25-28October95

Tarragona (Spain)

Borsum R.B. B&W License renewal rule /maintenance rule - It's too soonto talk about changes

1993 PVP conf. p. 41-44 25-29 Jul1993

New-York

Bosnak R., M. Vagins,J. Vora

NRC U.S. NPAR approach tomanaging aging in operatingnuclear power plants

11 SMIRT p. 339-345

18-23 Aug1991

Tokyo,Japan

Bouat M., R. Godin EDF Cooz-A expert assessmentprogram

1993 PVP Conf. p. 45-49 25-29 Jul1993

New-York

Bros J., J. Perelli Programa Integral de Vigilanciade las Estructuras de Hormigónde Centrales Nucleares


Session25-06

14-16October98

Valladolid(Spain)

Brown J.A., G.A. Tice PacificNuclearSystemsInc.

Containment penetrations -Flexible metallic bellows: Testing,safety, life extension issues

Nuclear Engineeringand Design 145(1993), pp. 419-430

9 July1993

Brumovsky M. Nucl.ResearchInst. Rezplc

Comparison of western andeastern codes for reactorcomponents lifetime assessment

ASME/JSME Conf. PVP 316p. 77-83

23-27 Jul1995

Honolulu

Brumovsky M., B.Gueorguiev

IAEA IAEA Co-ordinated researchprogramme on "Assuringstructural integrity of reactorpressure vessels"


Paper7473

April 19-23, 1999

Tokyo,Japan

Brumovsky M., M.Erve, C. Faidy, P.E.Mac Donald, T.R.Mager, J. Pachner, Ph.Tipping

IAEA WG IAEA documents on assessmentand management of ageing ofmajor NPP components importantto safety

PLIM + PLEX 97(Praha)

plim+plex97 pp.277-291

8December

1997

Burkhart D. AB StatensAnläggningsprovning

Thermal fatigue cracking inSwedish BWR feed watersystems

AB StatensAnläggningsprovning

20October

1980

Bustard L.D. SandiaNationalLaboratories

Management Plan for the LightWater Reactor Plant LifetimeImprovement Program


17 June1991

336

Cabrera E., M. Borrás,J. Ortega

MIDAS. Un nuevo Sistema deAdquisición de Datos paraInspección de ComponentesCríticos



25-28October95

Tarragona (Spain)

Calmand A., H.Michoux, J. Leclere

Framatome, EDF, CEA

Gestion des connaissances:maintien des conaissances etcompétences; problématique;l'exemple RNR


10December

1996

Carey J., M. Campbell,R. Nickell

EPRI Recent EPRI Life cyclemanagement programdevelopments

PLIM + PLEX 97(Praha)

plim+plex97 pp. 11-20+slides

8December

1997

Caro R. Management of NPP RemainigService Life - Present and Futurein Spain

PLIM+PLEX 95 Page 44,Session 2Views onSafety &Regulation

27-30November95

Nice(France)

Castaño M.L., A.M.Lancha, D. Gómez,F.J. Sanz

Examen de los Álabes de laTurbina de Alta Presión de la CNde Almaraz



25-28October95

Tarragona (Spain)

Castaño M.L., F.Blázquez, D. Gómez,A. Lagares

Velocidad de Crecimiento deGrieta del Inconel 600Sensibilizado, Contaminado conAzufre, en Reactores Tipo PWR


Session25-08

14-16October98

Valladolid(Spain)

Chaouadi R. SCK-CEN Fracture Toughnessmeasuremnts in the transitionregime using small size samples

Small SpecimenTest Techniques,1998

ASTMSTP 1329

Cleurennec M., Y.Thebault, E. Abittan, C.Pages, P.A. Lhote, L.Randrianarivo

EDF Expertises et contrôles des bridesde barrières thermiques despompes primaires équipant lestranches REP 900 Mwe.

Fontevraud iVproceedings

pp.791-801

17Septembe

r 1999

Cochet B. Framatome Des outils performants etinnovants au service de la duréede vie des équipementsmécaniques

RGN RevueGénérale Nucléaire

Nov. 1999

Combes J. P., J. F.Dubois, R. Godin

EDF Le projet durée de vie descentrales nucléaires

RGN - 1993 - N°3 -Mai - Juin, pp. 175-178

1 May1993

Combes J.P., R. Godin EDF EDF Lifetime project 1993 PVP conf. p. 21-26 25-29 Jul1993

New-York

337

Conner T., K.L.Saunders, P.A. Penn,M. Gielow

BaltimoreG and E.Hopperand Ass.

Nuclear power plant licenserenewal. Age related degradationinspection program


Paper ? April 19-23, 1999

Tokyo,Japan

Coste J.F., S. Jumel,R. Borrelly

EDFINSA Lyon

Non destructive characterizationof RPV steels by thermoelectricalpower measurements


Paper7082

April 19-23, 1999

Tokyo,Japan

Couplet D., G. Maes,H. Sweerts

TEE,Electrabel,AIB-Vinçotte

Belgian Experience: SomeExamples of using TechnicalJustification for NDT systemqualification

ENIQ workshop onTechnicalJustification, Petten,1-2 April 1998

1 April1998

Couplet D., P. Simoens TEE,Electrabel

The Belgian Policy forqualification of UT System: Acombination of ASME XIRequirements and ENIQGuidelines

1st InternationalConference on NDEin Relation toStructural Integrityfor Nuclear andPressurisedComponents,Amsterdam, 20-22October 1998

20October

1998

Courcoux A. Framatome Diagnosis and prevention ofageing phenomena in mechanicalequipment

SVA-Vertiefungskurs CONF-9411187p.3,1-1-3,1-10

2-4 Nov1994

Wintherthur (Switzerland)

Cragg C., A.K. Ghosh,W. Heep, R. Judge,D.J. Naus, J. Pachner,C. Seni, T. Tai, V.Vydra

IAEAcoordintedresearchprogram(CRP)

A generic framework for ageingmanagement of concretecontainments buildings

PLIM+PLEX 97(Praha)

plim+plex97pp.377-388

10December

1997

Cranford E.L., A.W.Engel, A.K. Kundu

Westinghouse, TUElectric Co.

The Comanche peak steamelectric station thermal eventmonitoring system

ASME/JSME Conf. PVP 316pp. 173-181

23-27 Jul1995

Honolulu

Crutzen S., B. Hems-worth, K.Kussmaul, M. Davies, P. Lemaître,R. Hurst, U. von Erstorff

The European Networks: NESC,AMES, ENIQ

Paper 95/70/1935 1995 Several institutions in Europe, including the JointResearch Centre have capabilities to deal withseveral of the problems posed by the ageing ofstructural components and their structural integrityassessment. The paper presents the cooperativeprogrammes organised in networks by theseinstitutions,

338

Daoust P., D. Couplet TEE Alloy 600 - Belgian experience(copy of slides)

1994 EPRIWorkshop onPWSCC of Alloy 600in PWR’s, Tampa,Florida, November15-17, 1994

EPRI TR-105406

1 August1995

Daoust P., D. Couplet,R. Gerard

TEE Results of the Doel 1-2 BMI andSI Nozzle Inspections

1997 EPRIWorkshop onPWSCC of Alloy 600in PWR’s, Daytona,Florida, February 25-27, 1997

EPRI TR-109138-P1

1November

1997

Davies L.M., S.Crutzen, U. vonErstorff, D. Sycamore

LMDConsult.JRC-IAMDGXVII/C3

AEPLAF - European PlantLifetime Assessment Forum: AnAMES/EC-DG XVII Initiative

5th Int. Conf. On "MaterialIssues in Design, Manufacturingand Operation of NPPEquipment"

June 19-26, 1998

St.Peters-burg,Russia

The paper presents the terms of reference ofEPLAF, including members, objectives, prioritiesand methodology of work. Additionally, a first draftwill be presented as an action plan for co-operationon technical issues of embrittlement and annealingof WWER pressure vessels.

De Marneffe L., P.Simoens, L. Imschoot

TEE Prestress Behaviour in BelgianNPP Containments

WANO-PC/OECD-NEA workshop onprestress loss inNPP containments,Poitiers 25-26August, 1997

De Preneuf R., F.Crassous, E.Raimondo

Framatome Plant life management 10, Pacific basinnuclear conf.

pp. 614-618

20-25 Oct1996

Kobe(Japan)

De Smet M., M.Guyette

TEE Justifying Fatigue Induced bySevere Thermal StratificationTransients in Feed WaterSystems

ASME 97 PressureVessel and PipingConference,Orlando, Florida,PVP-Vol. 350,Fatigue andFracture: 1997, Vol.1, pp. 441-445

De Smet M., M.Guyette, W.D’Haeseleer

TEE Fatigue Monitoring 2nd BelgatomInternationalConference -Brussels, May 21-24, 1995

21 May1995

339

Debarberis L., B.Acosta, M. Beers, J.F.Coste, J.P. Massoud,M.G. Horsten, P.Kauppinen, J.Pitkänen, G. Dobmann,M. Kröning, J. Bros, M.Hutchings, A.Rogerson, D. GomezBriceno, F.J. Porosanz,D.H. Stegemann, J.C.Spanner

JRC-IAM,EDF, NRG,VTT,FraunhoferIZFP,ARCS,AEA,CIEMAT,PSI, EPRI

Ageing Material Evaluation andStudies by Non-DestructiveTechniques AMES-NDTConcerted Action


29 Nov,-1Dec, 1999

Luxembourg

Delgado J.A., J.Vanhoomissen

Inspección del Shroud enCentrales BWR



25-28October95

Tarragona (Spain)

Dobbeni D. Laborelec NDT Techniques for nuclearcomponents

2nd BelgatomInternationalConference -Brussels, May 21-24, 1995

21 May1995

Dubois J.F., B.Granger, J.C. Fournel

EDF Probabilistic study for steamgenerator tube maintenance

1993 press. Vess.And piping conf.

p. 105-112

25-29 Jul1993

Denver(US)

Dwivedy K.K. VirginiaPower

Use of plant operating history todefine transient loads


21-26 Jul1996

Montreal

Erve M. Siemens Overall concept for Maintenanceand Plant Life Management(OECD/NEA - InternationalWorkshop On Nuclear PowerPlant Life Management)


14 April1997

Erve M., G. Bartholomé Siemens Activities in the field of plant lifeevaluation, life extension andplant improvement


19 July1990

Erve M., G. Maussner,N. Wieling, E.Tenckhoff

Siemens/KWU

Aging assessment and plant lifemanagement

Kerntechnik p. 353-359

Dec. 1992

Erve M., W. Kastner Siemens/KWU

Aging management for pipeworkparticularly considering corrosiveeffects


pp. 221-236(German)

25-26 Jan1996

SalgitterGermany

340

Eussen G., R. Duclos,D. Thomas

TEE 1E qualification of operationalelectrical and I&C equipment

Nuclear EuropeWorldscan 9-10,1996

Fabry A. SCK-CEN Characterization by notched andprecracked charpy tests of the in-service degradation of reactorpressure vessel steel fracturetoughness


ASTMSTP 1329

Fabry A., E. van Walle,R. Gerard, J. Van deVelde, R. Chaouadi, M.Mc Gough, H. Kwech

SCK-CEN,TEE, PCIEnergyServices

Enhanced Surveillance ofNuclear Reactor PressureVessels

Materials ageing andcomponent Lifeextension, pp. 577-588

Fabry A., R. Chaouadi,J. Van de Velde

SCK-CEN Comparison of BR3 Surveillanceand Vessel Plates to theSurrogate plates representativeof the Yankee Rowe PWR Vessel

18th InternationalSymposium; Effectsof Radiation onMaterials, WestConshohocken,1999

ASTMSTP 1325

Faidy C., G. Chas, S.Bhandari, M.P. Valeta,R. Hurst, A. youtsos, P.Nevasmaa, W. Brocks,D. Lidbury, C. Wiesner

EDF,Framatome, CEA,JRC, VTT,GKSS,AEAT, TWI

BIMET: Structural Integrity of Bi-Metallic Components


29 Nov,-1Dec, 1999

Luxembourg

Fernández R.A, F.Ceballos, D.Fernández

Modernización del Sistema deControl del Reactor de SantaMaría de Garoña


Page 47,Session4-03

22-26October96

Santander (Spain)

Flórez A.M., J. Jiménez Inspección por Ultrasonidos delBarrilete en Reactor BWR



25-28October95

Tarragona (Spain)

Fournier I. EDF Evaluation du vieillissement réeldes zones sensibles du circuitprimaire des centrales nucléairespar un système de surveillanceen fatigue

5/SF2M-C. Revuede métallurgie.Cahiersd'informationstechniques

Ref. EDF:E1993H563218

1 January1993

Francia L. Development Project of a NuclearPower Residual LifetimeEvaluation System


Session A 15-17November1994

Tokyo(Japan)

Francia L. Policy and Methodology on NPPLifetime Management in Spain


Page 187 28-30September 1998

Vienna(Austria)

341

Francia L. Alargamiento de Vida yExplotación de CentralesNucleares

Revista de laSociedad NuclearEspañola

Page 33 March1994

Madrid

Francia L. Aspects of Lifetime Managementin Spanish Nuclear Power Plants

Revista de laSociedad NuclearEspañola(International Issue)

Page 25 December1995

Madrid

Francia L., M. Regano Unesa,Nuclenor

Approach and status of NPPlifetime management in Spain

Plim + Plex 99 3-5November1999

Madrid,Spain

Fuchs R., Jan Stejskal KKWLeibstadtBKW FMBEnergie

Ageing Management and PLEXin Swiss Nuklear Power Plants

The paper presents the activities related to agingmanagement in Switzerland. Bases for agingmanagement are presented, with examples of Plantspecific component evaluation procedures.

García -Mazario M., C. Maffiotte,A.M. Lancha M. Hernandez-mayoral

Efecto del Azufre en las Capasde Óxido de las Aleaciones BaseNiquel Ensayadas en Ambientesde Secundario


,Session23-02

25-28October95

Tarragona (Spain)

Gauchet J.P., F.Munoz, E. Molinie, B.Bussy, N. Gillet

EDF,Framatome

Steam Generator InternalsDesign Review Part I: BundleWrapper Studies / Status andPerspective

SMIRT-15 August 15-20, 1999

Seoul,Korea

Gerard R. TEE Life Management of Doel 1-2Reactor Pressure Vessels


Paper7445

April 19-23, 1999

Tokyo,Japan

Gerard R. TEE Belgian Approach to thesurveillance of the ReactorPressure Vessel SteelEmbrittlement

PLIM+PLEX 95,Nice, November 27-30, 1995, pp. 424-434

27November

1995

Gerard R., A. Fabry, E.van Walle, J. Van deVelde, R. Chaouadi, M.Mc Gough, H. Kwech

TEE, SCK-CEN

In-Service Embrittlement of thepressure vessel welds at the DoelI and II Nuclear Power Plants

17th InternationalSymposium; Effectsof Radiation onMaterials,Philadelphia, 1996

ASTMSTP 1271

Gerard R., J. Van deVelde

TEE, SCK-CEN

The Life-Limiting aspects of theReactor Pressure vessels

BNS-SFEN Meetingon Life extension ofnuclear installations,Brussels, June 3-4,1992

3 June1992

342

Gerard R., R. deWouters, A. Fabry, J.Van de Velde

TEE, SCK-CEN

Advanced Methods for assessingRPV material degradation

2nd BelgatomInternationalConference -Brussels, May 21-24, 1995

21 May1995

Gillemot F., P. Kovacs AtomicEnergyResearchInstitute

NPP lifetime management inHungary

Nuclear Energy n°6 NE n°6,pp. 421-425

1Dcember

1996

Godin R. EDF Aging management ASME/JSME Conf. PVP-316p.113-128

23-27 Jul1995

Honolulu

Godin R. ATS(Finland)

Nuclear plant life management -the view from France

ATSYET ISSN:0356-0473 p.6-8

16 June1995

Godin R. EDF/SPT-BNS

Life extension of nuclearinstallations

BNS/SFEN Seminarproceedings atBrussels (B)

3 June1992

Gómez-Briceño D., F.Blázquez, F.Hernandez

Iniciación de grietas en Inconel600 en condiciones del CircuitoPrimario de los Reactores TipoPWR



25-28October95

Tarragona (Spain)

Gómez-Briceño D., F.Blázquez, F.Hernandez

Corrosión Bajo Tensión enComponentes de Inconel 600 enReactores Tipo PWR



5-7November97

LaCoruña(Spain)

Gosselin S.R. EPRI EPRI's new in-service pipeinspection program

Nuclear News p. 42 Nov. 1997 Risk-based ISI programs developed in-house willhelp reduce ISI examination scope and costs, andmaintain high safety standards.

Grandemange J.M.Y. Meyzaud

Framatome Outils d'évaluation duvieillissement en service descomposants lourds

Cours SVA/ASPEA 1 April1990

Brugg-Windisch, Switzerland

Grandemange J.M., C.Pichon

FramatomeEDF

Les modes de dégradation descomposants du circuit primairedes REP

Journée SFEN "Sûreté desinstallations anciennes"

8 déc.1998

Paris

343

Gueorguiev B., V.Lyssakov

IAEA IAEA activities in the field ofNuclear Power Plant LifeManagement (OECD/NEA -International Workshop OnNuclear Power Plant LifeManagement)


14 April1997

Gueorguiev B., V.Lyssakov, L.M. Davies

IAEA IAEA Activities in the Field ofNPP Life Management

IAEA

Guinot M. Framatome Remplacements de générateursde vapeur. La recherche d'unoptimum par Framatome


10December

1996

Guyette M. TEE Prediction of Fluid Temperaturefrom measurements of outsidewall temperatures in pipes

ASME Journal ofPressure VesselTechnology Vol.116, May 1994, pp.179-187

1 May1994

Guyette M., C. Leblois Belgian Experience with thestratification in the surge line

PLEX 91, Berlin,December 2-3, 1991

2December

1991

Guyette M., M. DeSmet

TEE Expected and surprising thermalstratification Phenomenaobserved in the Belgian PWRFeed Water Systems

Meeting onExperience withThermal Fatigue inLWR Piping causedby mixing andstratification, Paris,8-10 June 1998

NEA/CSNI/R(98)8

18December

1998


TEE Thermo-Mechanical AnalysisMethods for the conception andthe Follow up of componentssubjected to thermal stratificationtransients

Meeting onExperience withThermal Fatigue inLWR Piping causedby mixing andstratification, Paris,8-10 June 1998

NEA/CSNI/R(98)8

18December

1998


TEE Inverse and Direct TransferFunctions for Fatigue Follow upof Piping Systems submitted tostratification

Nuclear Engineeringand Design, 153(1995), pp.183-195

344

Hagberth R. OKG From the Oldest to the Youngest- Oskarshamn 1 - ModernizationStrategy (copy of slides)


2 October1997

Hamaguchi S. KansalElectricPower Co.(KEPGO)

Plant Life Management Studiesfor Mihama unit 1 (copy of slides)

Presentation 1Septembe

r 1996

Harrison L. DOE Plant Lifetime ImprovementProgram (copy of slides)


21 May1991

Henshaw J., P. Scott,R. Killian, O. Hietenan,R. Ahlstrand

AEA,FramatomeSiemens,IVO

Assessments of modellingrequirements to follow ageingphenomena: Modage


29 Nov,-1Dec, 1999

Luxembourg

Hernández F., M.S.García, D. Gómez

Efectos de las Especies conAzufre en el Secundario de losGeneradores de Vapor



25-28October95

Tarragona (Spain)

Hevia F. Mantenimiento para la Gestión deVida


Page 53,Session4-07

22-26October96

Santander (Spain)

Hevia F. Methodologies for MaintenanceEvaluation and Improvement

PLIM +PLEX 95 Page 513,PosterSession 1

27-30November95

Nice(France)

Hevia F., D.F. Lehnert,F.E. Gregor

Gestion de vida; aspectostecnicos y reguladores en EE.UU.

23rd annual meetingSociedad nuclearEspanola


5-7november1997

Lacoruna(Spain)

The purpose of the paper is to present an overviewof the current situation regarding actuations in thefield of Life Cycle Management and Life Extensionin the US.

Hevia Rupérez F. EmpresariosAgrupados

NPP Life ManagementProgrammes

InternationalCooperation forNuclearDevelopment, Kiev,July 1995

1 July1995

345

Hevia Rupérez F. EmpresariosAgrupados

Spanish industry gets together ona residual lifetime evaluationsystem

Nuclear EngineeringInternational,December 1993, pp.34-35

1December

1993

Hienstorfer W.G. TÜVSuedwest

Implementation of the aging plantlife management concept in theNeckarwestheim PWR from theangle of the reviewing expert


pp. 67-81(German)

25-26 Jan1996

SalzgitterGermany

Hill E.S., W.D. Midgett NUS Defining a coordinated approachto maintenance ruleimplementation and licenserenewal

1993 PVP Conf. p. 27-32 25-29 Jul1993

New-York

Hinkle W.D. YankeeatomicElectric Co.

Yankee nuclear power stationlicense renewal assessment

ANS annual meeting p. 290-291

7-12 Jun1992

Boston

Huettner C., F. Binder,G. Kroiss

TÜVMünchen

Aging management based on in-service fatigue analyses


pp. 221-236(German)

25-26 Jan1996

SalgitterGermany

Hutin J.P. EDF Plant Life Management: AnIntegral Part of Operation andMaintenance Policy andStrategies

Hutin J.P. EDF Stratégie de maintenance et deremplacement des groscomposants des centrales REP:l'optimisation des décisions


10December

1996

Ibanez R.L., C.E.Meyer

W Benefits derived from the use ofthe wog technical reports in lifecycle management

24 Annual meetingof the SpanishNuclear Society

pp. 37-44 14-16October

1998

Valladolid(Spain)

Ito D., M. Aoki, S.Maeda

TEPCHitachiToshiba

A feasibility study on nuclearreactor vessel replacement


Paper7292

April 19-23, 1999

Tokyo,Japan

Jacquot J.P., A.Dubreuil-Chambardel,A. Lannoy, B. Monnier

EDF The OMF project: Developmentof methods and tools forreliability-centered maintenance

1993 PVP Conf. p. 83-89 25-29 Jul1993

New-York

Jacquot J.P., C.Meuwisse, J.P.Cailleaux

EDF Improvement of the preventivemaintenance program of anemergency diesel generator set

ASME/JSME Conf. PVP 316p. 85-89

23-27 Jul1995

Honolulu

346

Jeong I.S., C.H. Jang KEPR Major Components LifeEvaluation for Nuclear PLIMFeasibility Study


Seoul,Korea

Jong-seog, K.Ill-seok, J.

KEPCO IAEA NPP Lifetime Management: experience of PLIM in Korea,presented at Vienna , 21-23 dec.1998

IAEA NPP LifeManagementDocument

21December

1998

Joosten J. Connect -USA LLC

Western European plant ageingissues: an overview

Nuclear EngineeringInternational, June1997, pp. 23-25

1 June1997

Jousselin A., R.Chevalier

EDF Diagnosis of faults in EDF powerplants: From monitoring todiagnosis

1994 PVP Conf. p. 63-67 19-23 Jun1994

Minneapolis

Kanshansky, Vermaut Westinghoue/EDF

Equipment lifetime monitors forFessenheim 1 - Program status

Slides slides 20November

1996

Karzov G., B. Timofeev PrometeyInstitute

Main mechanisms of materialproperties degradation underreactor pressure vessel operatingconditions


Paper7240

April 19-23, 1999

Tokyo,Japan

Kato N. , S. Kataoka,H. Sasajima, T. Otsuka

JAPEIC JAPEIC’s activity on aging issuesof reactor (pressure) vessel andits internals


Paper7338

April 19-23, 1999

Tokyo,Japan

Kaushansky M., M.Vermaut, P. Monette

Westinghouse

Compilation of aging degradationphenomena that affect criticalcomponents (internationalsurvey)

PLIM+PLEX 97(Praha)

PLIM+PLEX 97 -pp. 261-270

8December

1997

Kaushansky, Vermaut Westinghouse

NOK PLEX Phase 2 (copy ofslides)

EdF/WestinghousePLEX 27, 20/11/96

20November

1996

Kim I.S., J.S. Lee Korea Adv.Inst. Of Sc.And Techn.

Effect of intercritical annealingtreatment on the fracturetoughness of SA106Gr.C pipingsteel


Paper7093

April 19-23, 1999

Tokyo,Japan

Kim J. s., I. s. Jeong KoreaElectricPowerCooperation

Meeting Document: to beincluded in IAEA NPP LifeManagement Document

Consultancy meetingon NPP Plant LifeManagement,Vienna, 21-23December 1998

21December

1998

347

Kim J.S., I.S. Jeong,T.E. Jin

KEPRI,KPEC

Lifetime Management Study onKori Unit 1


Seoul,Korea

Kobayashi Y., T.Taguchi, I. Akiba, H.Mizoguchi, S. Yoshida,F. aida

Showa E.Toshiba

Combined thermal and radiationdegradation and diagnosis ofXLPE low voltage cable


Paper7166

April 19-23, 1999

Tokyo,Japan

Koopman R.B.C. EPZ Plant Life Management andLicense Renewal (OECD/NEA -International Workshop OnNuclear Power Plant LifeManagement)


14 April1997

Kosugiyama S., K.Takeuchi

TEPCJAPC

Plant life management activitiesof LWR plants in Japan


Paper7290

April 19-23, 1999

Tokyo,Japan

Kovacs P. PaksNuclearPowerPlant(MVM)

PAKS approach of PLIM(OECD/NEA - InternationalWorkshop On Nuclear PowerPlant Life Management)


14 April1997

Kovan D. Life after 40 Nuclear EngineeringInternational, pp. 26

Koyama K. JAPEIC Plant Aging, Service LifeManagement and Current Topicsin Japan (Copy of slides + text)

21st meeting ofPWG-3 on Integrityof Components andStructures - 3rdmeeting on Integrityof Metal Comp. andStru., Brussels, 16-18 June 1998

16 June1998

Koyama M. JAPEIC Current status of the approachesand measures for the ageingNPPs in Japan

20st meeting ofPWG-3 on Integrityof Components andStructures - 2ndmeeting on Integrityof Metal Comp. andStru., Paris, 12-13June 1997

12 June1997

348

Koyama M. JAPEIC Current Status of the Approachesand Measures for the AgingNuclear Power Plants in Japan(OECD/NEA - InternationalWorkshop On Nuclear PowerPlant Life Management)


14 April1997

Koyama, Masakuni,Tadahashi, Takao,Mimaki, Hidehito,Ando, Masami,Kanazawa, Yasushi,Yoshitsugu, Mishima

JAREICMitsubishiHitachiToshibaTokyoUniv.

Current status of PLEX project atJAPEIC

ASME Conf. p. 107-113

21-26 Jul1996

Montreal

Krett V. IAEA activities on nuclear powerplant ageing and life extension.Operability of nuclear systems innormal and adverseenvironments

SFEN Congress Ref. EDF:E1990E133619

1 January1989

Kruger D. A. InstrumentationServicesInc.

Optimize predictive maintenancefor rotating and electricalequipment

HydrocarbonProcessing, January1998

1 January1998

Kussmaul K., E. Roos,S. Burnay, F. Michel,K. Knips, C.J. Bolton,D.W. Twidale, J.M.Humbert, N.Mermilloid, P.Petrequin, L.Debarberis, U. vonEstorff, F. Sevini, M.Erve, A. Nink, W.Michel, J.C. Cano, J.Eibl, R. Rintamaa

MPA, GRS,Magnox,CEA, JRC-IAM,Siemens,Tecnatom,Univ.Karlsruhe,VTT

Integrity Assessment of AgeingComponents. Concerted ActionINTACT.


29 Nov,-1Dec, 1999

Luxembourg

Laaksonen J. STUK Giving VVERs a new lifetime Nucl. Eng. Int. Vol 43, No528 p.32-4, 36

July 1998 Finland

Lafaille J. P., J. C.Hennart

TEE Mechanical DegradationProcesses: The BelgianExperience

Conference ENC 98,Nice

349

Lai H.C. Component life assessmentmethods for nuclear power plant

2nd Int. ColloqiumAgeing of materialsand methods for theassessment andextension of lifetimesof eng. Plant

5/0308-0161Elsevier

1 January1994

Lancha A.M, M. Serrano, J.Lapeña, D. Goméz-Briceño

Análisis de Fallo del Eje de unaBomba de Circulación de Aguade Rio de la CN Ascó I



5-7November97

LaCoruña(Spain)

Lapeña J., F.J.Perosanz, M. Gachuz

Recostrucción por "Stud Welding"de Probetas Charpy-VPertenecientes a Cápsulas deVigilancia


Session10-02

14-16October98

Valladolid(Spain)

Le Doare C. IPSN Sûreté des installationsanciennes, La démarche deréexamen de sûreté desréacteurs à eau sous pression

SFEN conference Dec. 8,1998

Paris The paper presents the point of view of the Frenchsafety authority concerning periodic safetyreevaluation of NPPs.

Leblois C. TEE Optimisation of costs for the Doel3 Steam Generator Replacement

IAEA Specialists'Meeting on SteamGenerator problemsand replacement,Madrid, 13-16December 93

13December

1993

Leblois C., J. Berthe,D. Mallieu

TEE,Electrabel

Steam Generators Managementin Belgium

AIM 9/93 1Septembe

r 1993

Lee N.H., J.S. Kim KOPEC Aging and life extension ofpressurized water reactor steelcontainment, KORI Unit 1

ASME Conf. pp. 129-135

21-26 Jul1996

Montreal

Leggatt R.H., E. Keim,P.J. Bouchard, H.Keinanen, M. Clerge,F. Rustichelli, D.J.Smith, V. Schulze, A.G.Youtsos, C.P. Millward,S.K. Bate

TWI,Siemens,Nucl. Elec.,VTT, IS,Univ.Ancona,Univ.Bristol,IWK, CECJRC,Magnox,AEA.

Variations of residual stresses inaged components


29 Nov,-1Dec, 1999

Luxembourg

350

Lemaitre P., G. VanGoethem

EuropeanCommission

EU research activities in the fieldof plant life extension andmanagement: Achievements ofFP-4 and prospects for FP-5

2000 ASME Pressure Vesseland Piping Conference

July 23-27, 2000

Seattle,USA

Lietard J.P., D Couplet,P. Simoens, H.Sweerts, L. Hernandez

TEE,Electrabel,Intercontrole

Belgian Experience in ReactorPressure Vessel In-ServiceInspection

EPRI 96, ReactorPressure VesselInspectionConference, OlympicValley USA, 4-6June 1996

4 June1996

Lofaro R., W. Gunther,M. Subudhi, J. Taylor

Brookhaven NationalLaboratory

Nuclear Plant Aging Study ofResidual Heat Removal Systems

Nuclear Engineeringand Design 118(1990) pp. 375-380

19 July1989

Lofaro R.J. Brookhaven Nat.Labs.

Aging mechanisms and failuremodes in nuclear power plantcontainment spray systems

1993 PVP conf. p. 149-160

25-29 Jul1993

New-York

López N., J. García, M.Rodríguez

Sistema para el Diagnóstico demalfunciones de Componentes



25-28October95

Tarragona (Spain)

Marcelles I., M.Aguado, J.J. Latova,G.L. Stevens

Gestión del Envejecimiento enBase a Monitorización de Fatiga


Session10-04

14-16October98

Valladolid(Spain)

Martín J. Development of a Nuclear PowerResidual Lifetime EvaluationSystem

PLIM +PLEX 95 Page 214,Session 4Plant LifeProgrammes

27-30November95

Nice(France)

Martín J., J.L. Jimeno,P. García

Plan de Gestión de Vida de laCentral Nuclear José Cabrera


Page 48,Session4-04

22-26October96

Santander (Spain)

Martínez P.J., J.B.Blázquez

Mantenimiento Predictivo deCables en Plantas Nclearesmediante Técnicas de Análisis deRuido



25-28October95

Tarragona (Spain)

Martorell S., A.Sánchez, V. Serradell

Modelo de EnvejecimientoIncorporando el Efecto delMantenimiento y de lasCondiciones de Servicio


Session25-01

14-16October98

Valladolid(Spain)

351

Masson S. H., J. P.Hutin, F. Tetreau

EDF,Framatome

Prise en compte industrielle durisque d’obsolescence


10December

1996

Massoud J.P., J.F.Coste, J.M. Leborgne,D. Aiguier, P. Viral, G.Bedzikian, H. Churier-Bossennec

EDF Thermal aging of PWR duplexstainless steel components -development of a thermoelectricaltechnique as a non-destructiveevaluation method of aging


Paper7240

April 19-23, 1999

Tokyo,Japan

Mathonet J., J. M.Cherasse, L. C.Leblois, M. Guyette

TEE Belgian Methodology forEvaluation of Erosion-CorrosionPhenomena in units 1, 2, 3 & 4 ofthe Doel Nuclear Power Plant

ASME PressureVessel and PipingConference, July 23-27, Hawaï

23 July1995

Matsubara S., Y.Yokono, Y. Nagano, T.Imanaka, Y.Kawaguchi, H.Mitsuda, Y. Okano

ND CoInst. OfNucl.SafetySyst. Inc.

Evaluation of thermal aging ofduplex stainless steel by meansof sound velocity measurement inthe fine region using phaseinterference technique


Paper7132

April 19-23, 1999

Tokyo,Japan

Meauzoone L. EDF Contrôle commande et durée devie


Paris The paper presents the actions taken andenvisaged concerning I&C ageing

Meyer L. C., H.L.Magleby

IdahoNationalEngineeringLaboratory

Nuclear Plant Aging Research onthe high-pressure injectionsystem

Nuclear Engineeringand Design 118(1990) pp. 389-398

12Septembe

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Meyzaud Y., P. Soulat FramatomeCEA

In-serice aging of pressurizedwater reactor steam supplysystem materials

RGN Int. Edition,Vol.A

pp.5-9 July 1996

Michel W. Siemens Prognosis on the aging of cables Plim + Plex Conf. 08-10December1997

Prague / Czech Republic

Michel W. Siemens -KWU

Prognosis on the ageing oncables

PLIM+PLEX 97(Praha)

plim+plex97 pp.377-387

10December

1997

Miller A. OECD/NEA

NEA Nuclear Safety DivisionActivities (OECD/NEA -International Workshop OnNuclear Power Plant LifeManagement)


14 July1997

352

Miller A.G. , J.P.Clausner, B. Kaufer

OECD/NEA

Activities on component reliabilityunder the OECD NEA

ICONE 5,conference onNuclear engineeringat Nice, France

ICONE 5 -2492

26 May1997

Miller A.G., L.M. Smith OECD,BritishEnergyGeneration

Activities of the OECD NuclearEnergy Agency in the Area ofConcrete Containment Ageing


Seoul,Korea

Mimaki, Hidehito,Kanasaki, Hiroshi,Suzuki, Isao, Koyama,Masakuni, Akiyama,Mamoru, Mishima,Yoshitsugu, Okuba,Tadatsune

MitsubishiJAPEICTokyoUniv.SophiaUniv.

Material aging research programfor PWR plants

ASME Conf. p. 97-105 21-26 Jul1996

Montreal

Munoz F., J.P.Gauchet, M. Stindel, C.Dabere, A. Pradel

EDF,Framatome

SG Internal Design Review PartIII: Corrective and PreventiveMaintenance and in-serviceSurveillance of Steam GeneratorBundle Wrappers


Seoul,Korea

Muri W. ABB Improvements & Benefits by HeatExchanger Retrofits

ABB Nuclear ServiceSymposium, June16-18, 1998 inStockholm

16 June1998

Naisse J.C. TEE I&C Modernization Strategy 2nd BelgatomInternationalConference -Brussels, May 21-24, 1995

21 May1995

Nakamura T., K.Bunno, S. Masamori

Kansai El.Power Co

Plant life management studies forjapanese PWR plant

ASME/JSME Jointpress. Vess. AndPiping Conf.

PVP 360p. 427-432

20 June1995

SanDiegoUSA

Nakamuta Y., T.Miyoshi, E. O’shima

JAPEICTokyo Inst.of techn.

In-operation inspectiontechnology development<Development of the degradationprediction technique>


Paper7154

April 19-23, 1999

Tokyo,Japan

Naus D., C.B. Oland,B. Ellingwood, W.E.Norris, H.L. Graves

NRC Management of aging of nuclearpower plant containmentstructures

4th Int. Conf. Oneng. Struct. IntegrityAssessment

ORNL/CP-96731(9p)

31 Dec1998

Washington

353

Naus D.J., B.R.Ellingwood, J.L.Cherry, N.C. Chokshi,J.F. Costello

ORNL,JohnHopkinsUniv.Sandia, USNRC

Nuclear Power PlantContainment Pressure BoundaryAging Research


Seoul,Korea

Naus D.J., C.B. Oland,B. Ellingwood, Y. Mori,E.G. Arndt

ORNLJohnHopkinsUniv. NRC

Towards assuring the continuedperformance of safety-relatedconcrete structures in nuclearpower plants

1993 PVP conf. p. 121-137

25-29 Jul1993

New-York

Naus D.J., C.B. Oland,B.R. Ellingwood, H.L.Graves, W.E. Norris

ORNLUniv.BaltimoreNRC

Aging management ofcontainment structures in nuclearpower plants

3rd int. Conf. Oncontainment designand operation

p.17 19-21 Oct1994

Toronto(Canada)

Naus D.J., C.B. Oland, B.R.Ellingwood, W.E. Norris, H.L.Graves

Factors related to agingmanagement of Nuclear PowerPlant containment structures

ASME Conf. Ref. EDF:E1999E100253

janv-98

Neils G.H. NorthernStatesPowerCompany

US Utilitiy Perspective on NuclearPlant Life Extension

11th InternationalSMIRT (StructuralMechanics InReactor Technology)Conference, August18-23, 1991, Tokyo

18 August1991

Nishida Y., K. Itagaki,Y. Yamaji, M. Konishi,S. Suzuki, T.Yamamoto

Kansai E.Hokkaido EShikoku E.Kyushu E.JAPCMitsubishi

Non-destructive diagnosistechnique for aging of cablesused at nuclear power plants


Paper7274

April 19-23, 1999

Tokyo,Japan

Noel R. EDF Durée de vie des centralesnucléaires REP

La TechniqueModerne, mai - Juin,1987, pp.27-35

1 May1987

Osaki K., Y. Watanabe,Y. Kitajima, H. Hattori,Y. Uhara, T. Miyoshi,E. O'shima

Toshiba,JAPEICTokyo Inst.Of Techn.

Development of degradationprediction technology for rotatingmachines


Paper7200

April 19-23, 1999

Tokyo,Japan

354

Otsuka T. JAPEIC Current Status of LifeManagement Policies for NuclearPower Plants in Japan (not all)

Specialist Meetingon Strategies andPolicies for NuclearPower Plant LifeManagement,Vienna, 28-30September 1998

28Septembe

r 1998

Pachner J. IAEA IAEA Products on the Evaluationand management of safetyaspects of NPP ageing


17 June1991

Pachner J. IAEA Systematic Ageing ManagementProcess: A Key Element for LongTerm Safety, Reliability andEconomy of Nuclear PowerPlants


Seoul,Korea

Palomo J. Iberdrola Contribution of modern plantmanagement to the improvementof safety culture


29 Nov,-1Dec, 1999

Luxembourg

Park J.S., H.Y. Roh,C.M. Lee, T.E Jin

KPEC The Development of ReactorPressure Vessel ThermalAnnealing Evaluation Program forNPP Lifetime Management


Seoul,Korea

Parshley P. C., D. F.Grosser, D. A. Roulett

ShearsonLehmanBrothers

Are Older Nuclear Plants StillEconomic

Lehman BrothersUtilities ResearchConference, Vol. 2,N°21

27 May1992

Petit R. EDF Life management andrefurbishment of nuclear units inFrance and in OECD countries


Paper7486

April 19-23, 1999

Tokyo,Japan

Pikelny B.V., M.H.Sanwarwalla

ComEdSargent &Lundy LLC

Reduction in surveillancerequirements for Limitorque valveoperators


21-26 Jul1996

Montreal

Pin T., D. Grypczynski FramatomeNuclearservices

5/8" Baffle bolt replacement Sociedad nuclearEspanola - 24threunion annual: 14-16/10/98

14October

1998

355

Pironet G., A. Heuzé,O. Goltrant, R. Cauvin

Electrabel,Framatome, EDF

Expertise des vis de liaisoncloison-renfort de CNT1

EDF 2 June1998

Poudroux G., N. Gillet,J.P. Gauchet, M. Petit,F. Munoz

FramatomeEDF

SG Internals Design Review PartII: SG Wrapper Drop SafetyStudies


Seoul,Korea

Pucak J.L., E.M. Brown ABB/CE Use of ABB advant power forlarge scale instrumentation &controls replacements in nuclearpower plants


Paper7458

April 19-23, 1999

Tokyo,Japan

PWG3 OECD/NEA/CNRA

Regulatory aspects on ageingreactors

OECD/NEA 18 May1998

Ramos M.A. Vigilancia del Envejecimiento deCables en la Central Nuclear deDukovany



5-7November97

LaCoruña(Spain)

Regan C. Nuclear power plant genericaging lessons learned (GALL)

SMIRT Ref. EDF:E1997E100675

1 January1997

Regano M. Nuclenor Spanish Approaches to PLIM oneconomic and regulatory aspects


14 April1997

Regaño M. Life Cycle Management - AFeasible and NecessaryProgramme

PLIM +PLEX 95 Page 140,Session 3UtilityPerspectives &Strategies

27-30November95

Nice(France)

Regaño M., L. Francia El Atractivo del Alargamiento deVida de las Centrales NuclearesEspañolas



5-7November97

LaCoruña(Spain)

Rinckel M.A. FramatomeTechnol.

Reactor pressure vessel integrityprogram

6th symp. On currentissues related toNPP structures:operation,maintenance andcost reduction

NuclearEng. AndDesign

4-6 Dec1996

May 1998,p. 17-39

Raleigh(US)

Rorive P., J. Berthe, J.P. Lafaille, G. Eussen

TEE,Electrabel

Nuclear Power Plant LifeManagement

Conference AIM,Liège, October 1997

356

Ross D.F. NRC Performance measures for agingof nuclear power plants

1993 PVP conf. p. 1-7 25-29 Jul1993

New-York

Sahgal S. Beznau Ageing Management in NuclearPower Plant Beznau


Seoul,Korea

Sanwarwalla M.H. Sargent &Lundy

Cable life extension for plantlicense renewal

1993 PVP conf. p.139-148 25-29 Jul1993

New-York

Sanwarwalla M.H., R.J.Weinacht

Sargent &Lundy,CalvertCliffs

Aging management throughcondition monitoring of ASCosolenoid valves and NAMCO limitswitches

ASME/JSME Conf. PVP 316,p.51-57

23-27 Jul1995

Honolulu

Savoldelli D., C. Faidy EDF Transient monitoring experiencein French PWR units

ASME Conf. pp. 89-95 21-26 Jul1996

Montreal

Schlaseman C.S., B.M.Tilden

MPRAssoc.BaltimoreGas andElectric Co.

A commodity approach to agingmanagement review of supportsfor license renewal

ICÔNE-4 Vol.5p. 197-203

10-13 Mar1996

NewOrleans

Schuler M. BCCN Le vieillissement des cuves SFEN conference Dec. 8,1998

Paris The paper presents the point of view of the FrenchSafety Authority on the RPV ageing.

Schulz H. GRS Köln Challenges in the areas ofmaterials ageing and plantmodernisation


29 Nov,-1Dec, 1999

Luxembourg

Schulz H. Gesellschaft fürReaktorSicherheit(GRS)

The evaluation of componentlifetime on the basis of operatingexperience


15January

1990

Scibetta M., R.Chaouadi

SCK-CEN Fracture Toughness DerivedFrom small circumferentiallycracked bars


ASTMSTP 1329

Seddon J.W., D.Goodison, E.M. Pape

Safety re-evaluation of ageingnuclear power plant in the UK

11th SMIRT 1 January1991

Shah V.N., U.P. Sinha,A.G. Ware

Idaho Nat.Eng. Lab.

Aging management of major lightwater components

Aging research Inf.Conf.

Nureg/CP-0122-Vol.1 p.250-274

24-27 Mar1992

Rockville

Shibata K. JAERI Progress of LWR StructuralSafety Research at JAERI

357

Shibata K. JAERI Research Program on Integrity ofAged LWR Components at JapanAtomic Energy Research Institute(copy of slides)


16 June1998

Shibata K. JAERI Overview of Aging and StructuralIntegrity Research Program atJapan Atomic Energy ResearchInstitute (copy of slides)

2nd meeting onIntegrity of MetalComponents andStructures, June1997

1 June1997

Sievers J. Gesellschaft fürReaktorSicherheit(GRS)

Status on Reactor PressureVessel Pressurized ThermalShock International ComparativeAssessment Study


16 June1998

Sighicelli S., M.Sabaton

EDF Contrôle du vieillissement, suivi,surveillance des matériels descentrales nucléaires


10December

1996

Simola K. Experience based ageinganalysis of NPP protectionautomation

ANS Congress Ref. EDF:E1999E100166

1 January1998

Somville P. TEE Comparison between Repair andReplacement Strategies for theDoel 3 Steam Generators

EPRI 90, Long-Range StrategicSteam GeneratorManagementworkshop, Denver,Colorado, July 11-13, 1990

11 July1990

358

Somville P., P.Hernalsteen, R.Houben

TEE,Electrabel,Laborelec

Steam Generator Replacementas a part of a general problemmanagement process

Operatingexperience withSteam Generators,Papers presented atCSNI/UNIPEDESpecialist Meeting,16-20 September1991, Brussels

16Septembe

r 1991

Staudinger D.K., W.R.Gray

B & W The B and W owners grouplicense renewal program

1993 PVP Conf. p. 9-14 25-29 Jul1993

New-York

Steiner E.K. Kernkraftw,Grafenrheinfeld

Strategies and action for plant lifemonitoring in German nuclearpower plants

JK’96 pp. 79-103(German)

21-23 May1996

Mannheim

Streicher V.J, U.Kunze, G. Engl

Application of monitoring systemsand inservice inspections inVVER plants

Plim+plex 1997 p. 187-196

Dec. 1997 Prague (Czech Republic)

Sung-Yull H. KoreaElectricPowerCooperation R&DCenter

Current Status of NPP AgingManagement, PerofrmanceImprovement, Life TimeExtension in Korea


17 June1991

Suzuki M., K. Onizawa,Y. Nishiyama, N. Ebine

JAERI Recent Progree on IrradiationEmbrittlement and RelatedReliability Studies on PressureVessel Steels

Proceeding of 19thKAIF/JAIF Seminaron Nuclear Industry(1997)

Takao T., N. Soneda,T. Sakai

CentralResearchInst. OfElectr.Power Ind.

Development of a support systemto make economic and technicalassessments for the issuesrelating to plant life extension

1994 PVP Conf. p. 43-50 19-23 Jun1994

Minneapolis

Tanarro A., A. García,P. Serrano, M. Tejo

Aplicación de los Sisemas Arraya la Inspección de internos deVasija



25-28October95

Tarragona (Spain)

Tang S.S., P.C.Riccardella, R. Dyle

Struct. Int.Assoc. Inc.SouthernNucl.Operations

The effect of reduction ininservice inspection on thereliability of boiling water reactorpressure vessels

ASME Conf. PVP-332,p. 49-57

21-26 Jul1996

Montreal

359

Tezuka K. , S.Kawamura

TokyoElectricPower Co.

Aging assessment and licenserenewals: Plant life managementfor the first stage boiling waterreactor

ICÔNE 4 Vol.5p. 271-275

10-13 Mar1996

NewOrleans

Tice D.R., P. Aaltonen,B. Rosborg, J.D.Atkinson, W. Dietzel,D.I. Swan, J. Lapena,S. McAllister

AEAT,VTT,Studsvik,SheffieldUniv.GKSS,Rolls-Royce,Ciemat,JRC-IAM

Evaluation of techniques forassessing corrosion cracking indissimilar metal weldments


29 Nov,-1Dec, 1999

Luxembourg

Toichi B., U. Shunsuke,S. Wataru, O. Asamu

Hitachi Aging management andpreventive maintenance fornuclear power plants

Hitachi Hyoron Vol.77,No4, p.301-306

1995 Japan

Tomkins B. AEATechnology

AEA Technology's presentation(OECD/NEA - InternationalWorkshop On Nuclear PowerPlant Life Management)


14 April1997

Torres V., L. Francia Gestión de Vida para unComponente Insustituible: ElEdificio de Contención


Session25-07, pp.63-85

14-16October98

Valladolid(Spain)

Tsujikura, Yonezo KansaiElectricPower

Plant life management studies fornuclear power plants

Thermal and nuclearpower

p.1078-1086

Se 1997 Japanese

Valibus L. EDF French perspective on lifemanagement of NPPs

EDF-EPN 4002-02/9604

20 March1996

The paper is dedicated to the presentation of EDF"Lifetime project".

Valibus L., J. Branchu EDFFramatome

Durée de vie des équipements ducircuit primaire principal

Conference SFEN "Le parcnucléaire: sa gestion dans ladurée"

10 et 11décembre1996

Paris After a general presentation of the plant lifemanagement, the main components are identifiedwith their potential degradation mechanisms.Developments dedicated to a better understandingof these degradation mechanisms and criteria arediscussed.

Valibus L., M. Sabaton EDF Durée de vie et vieillissement descentrales nucléaires à eaupressurisée d'EDF


Paris The paper is dedicated to the presentation of EDF"Lifetime project".

360

Van Goethem G. Euratom EU Research in Reactor Safety:Achievements of the 4th andprospects for the 5th EuratomFramework Programme

Eurocourse-99 on ReactorSafety Advanced nuclearReactor design and Safety

17-21 May1999

GRSGarching/munich

Van Walle E., M.Scibetta, M.J. Valo,H.W. Viehrig, H.Richter, T. Atkins, M.R.Wootton, E. Keim, L.Debarberis, M. Horsten

SCK-CEN,VTT, RCR,AEA,BNFL,Siemens,JRC-IAM,NRG

Reconstitution techniquesqualification & evaluation to studyageing phenomena of nuclearpressure vessel materials(RESQUE)


29 Nov,-1Dec, 1999

Luxembourg

Vançon D., Y.Meyzaud, P. Soulat

EDF,Framatome, CEA

Connaissance des phénomènesde vieillissement des matériauxutilisés dans les centralesnucléaires à eau sous pression


10December

1996

Various Lifetime Management andEvaluation of Plant, Structuresand Components

4th Int. Conf. On EngineeringStructural Integrity Assessment,Churchill College

22-24Sep. 1998

Cambridge, UK

Various RollsRoyce,Tecnatom,PreussenElektra,Belgatom,SAQ, AEA,OKG AB,TWI, EDF,JRC Petten

European Network of Risk-Informed In Service Inspection(RI-ISI) (EURIS)


29 Nov,-1Dec, 1999

Luxembourg

Various Prévision et estimation de ladurée de vie des structuresmécaniques

INSTRUC 4 23-24November1999

Paris,France

Various Le vieillissement des installationsnucléaires

Revue CONTRÔLEN°129, DSIN Paris

juin-99

Varley J. NEI Continued operation or prematureclsure: the billion dollar question

Nuclear EngineeringInternational

10December

1999

361

Vermaut M., P.Monette, T. Meyer, W.Turkowski

Westinghouse

Designing a LCM Program toFulfill Regulatory, Technical andEconomic Constraints (copy ofslides)

4th RegionalMeeting - NuclearEnergy in CentraEurope, Bled(Slovenia), 7-10September 1997

7Septembe

r 1997

Verón P. Fabricación de Defectos porCorrosión Bajo Tensión



5-7November97

LaCoruña(Spain)

Verón P. IGSCC en Tubos con Denting 24 Sociedad NuclearEspañola AnnualMeeting

Session10-03

14-16October98

Valladolid(Spain)

Vilemas J., E. Uspuras LithuanianEnergyInstitute

Ignalina's new safety programme Nuclear EngineeringInternational, June1997

1 June1997

Villain B., P. Pittner, H.Procaccia

EDF Probabilistic approaches to lifeprediction of nuclear plantstructural components

ASME Conf. PVP-332,p. 41-48

21-26 Jul1996

Montreal

Ward P.W., T.K.Shome

Stone &WebsterPublicServiceElectricand GasCo.

Containment liner evaluation andaging degradation managementprogram

ASME/JSME Conf. PVP-313-2 p.75-85

23-27 Jul1995

Honolulu

Watson P.C., B.Gooder

Ontariohydro

Ontario hydro nuclear Plant LifeManagement activities(OECD/NEA - InternationalWorkshop On Nuclear PowerPlant Life Management)


14 April1997

Watzinger H., M. Erve Siemens Plant Life management optimizedutilization of existing nuclearpower plants


Paper7505

April 19-23, 1999

Tokyo,Japan

Wiechmann K. Siemens Plant life management bycomponent replacement

IAEA Spec. Meet.On Power Plantcomp., maintenance,repair andreplacement for lifemanagement

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Wolter P. TÜVHannover/Sachsen

Aspects of fatigue monitoring inthe nuclear power plantsGrohnde (KWG) and Emsland(KKE)


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25-26 Jan1996

SalzgitterGermany

Yoshida K., S.Kataoka, A. Okamoto

Ishikawajima-HarimaHeavyIndustriesCo. (IHI)

Plant service continuanceapproach through periodicalexamination


13 May1991

Zaiss W. Neckarwestheim

Operating results from GermanPWRs shown with the example ofthe GKN reactor

JK'96 pp. 5-31(German)

21-23 May1996

Mannheim

Zdarek J. NRI VVER PHARE NDE Project 21st meeting ofPWG-3 on Integrityof Components andStructures - 3rdmeeting on Integrityof Metal Comp. andStru., Brussels, 16-18 June 1998

16 June1998

Zdarek J. NRI Reliability oriented PLIM OECD/NEA/PWG3 1 March1997

ANO-3's risk-informed ISIapplication

Nuclear News Page 78 July 1999