Strategy Maps 2007 for Ageing ManagementStrategy Maps 2007 for Ageing Management

Naoto Sekimura, Prof., Dr. Department of Quantum Engineering and Systems ScienceDepartment of Nuclear Engineering and Management

The University of Tokyo

IAEA IAEA PLiMPLiM 2007 Symposium 2007 Symposium Shanghai, China, October 15Shanghai, China, October 15--18, 200718, 2007

Technical cooperation among research organizations,Technical cooperation among research organizations,utilities and Universitutilities and Universitiesies in Japanin Japan-- Infrastructure for Ageing Management Infrastructure for Ageing Management --

-- Domestic and International Research Cooperation Domestic and International Research Cooperation ---- Database and Information Basis, Human Resources Database and Information Basis, Human Resources --

Technical Information Base

Comprehensive coordination

Industry � Technical development� Plant information, etc.

Academia

�Basic research �Scientific information, etc.

Government

� Safety research � Information on regulations,

etc.

Umbrella NetworkInformation Exchange

Other Industries

International Projects

A Coordinating Committee on Ageing ManagementA Coordinating Committee on Ageing Managementwas established with members from industries, utilities,

research organization and regulatory authorities.

2

R&D Roadmaps for Ageing ManagementR&D Roadmaps for Ageing Managementand Safe Long Term Operationand Safe Long Term Operation

developed by Atomic Energy Society of Japan in 2004developed by Atomic Energy Society of Japan in 2004--20052005

Database onRegulationProceduresIn OtherCountries

To Keep Safety and Reliability of Nuclear Power PlantsTo Keep Safety and Reliability of Nuclear Power Plantsfor Long Term Operationfor Long Term Operation

EstablishmentEstablishmentof Informationof Information

BasisBasisTechnicalTechnical

DevelopmentDevelopment Codes andCodes andStandardsStandards

SystematicSystematicMaintenanceMaintenance

SystematicAgeing

ManagementProgram

Database forDegradation of

Materials

RPVRadiationEmbrittle-lment

EvaluationEvaluationTechnologyTechnology

for Degradationfor Degradationof Componentsof Components

PerformanceIndex

SchemesTo ApplyNew

Techniques

Standardization ofStandardization ofAgeing ManagementAgeing Management

ProceduresProcedures

HumanResources

Risk-basedMaintenance

Optimization ofOptimization ofMaintenanceMaintenance

IASCC

1.1. 2.2. 3.3. 4.4.

3

Development of Development of Technical Information InfrastructureTechnical Information Infrastructure

Human Resources

Technical Information Infrastructure

SystematicMaintenanceImprovement

Development of Codes and Standards

Selection of Research Theme

Research result

-Maintenance Information -Case Maintenance -Operating Experience

Latest Information

Tacit Knowledge Tools for Education

Latest Information

Technical Evaluation Result

4

Safety Research & Development

Strategy of Safety Research and Related Strategy of Safety Research and Related Infrastructure for Ageing Management Infrastructure for Ageing Management

Ageing technical information databasePlant operation experience database

Technical information networkKnowledge management system

Regulatory System Infrastructure

BudgetSupport

FacilityInfrastructure

HumanResources

Codes and Standards

Maintenance Engineering of Plant Systems

Technological Information Infrasturcture

SafetyResearch

5

- Strategic training of expertise human resources

- international cooperation

R&D Roadmaps for Ageing ManagementR&D Roadmaps for Ageing Managementand Safe Long Term Operationand Safe Long Term Operation

developed by Atomic Energy Society of Japan in 2004developed by Atomic Energy Society of Japan in 2004--20052005

Database onRegulationProceduresIn OtherCountries

To Keep Safety and Reliability of Nuclear Power PlantsTo Keep Safety and Reliability of Nuclear Power Plantsfor Long Term Operationfor Long Term Operation

EstablishmentEstablishmentof Informationof Information

BasisBasisTechnicalTechnical

DevelopmentDevelopment Codes andCodes andStandardsStandards

SystematicSystematicMaintenanceMaintenance

SystematicAgeing

ManagementProgram

Database forDegradation of

Materials

RPVRadiationEmbrittle-lment

EvaluationEvaluationTechnologyTechnology

for Degradationfor Degradationof Componentsof Components

PerformanceIndex

SchemesTo ApplyNew

Techniques

Standardization ofStandardization ofAgeing ManagementAgeing Management

ProceduresProcedures

HumanResources

Risk-basedMaintenance

Optimization ofOptimization ofMaintenanceMaintenance

IASCC

1.1. 2.2. 3.3. 4.4.

6

Continuous Revision of Strategy MapsContinuous Revision of Strategy Maps forforAgeing Management by All the StakeholdersAgeing Management by All the Stakeholders

①①①①Introduction Scenario

②②②②Maps of Technical Issues

Strategy Maps for Ageing ManagementStrategy Maps for Ageing Management

Nuclear Safety Regulatory Standard Committee in NISA

Comments &Participation

Periodic Revision through Latest Knowledge

……

………

……

Publicity

• Needs• Targets• Milestones

• R&D projects• Information Basis• Hardware Resources• Human Resources

Safety Research Sub-Committee inthe Coordinating Committee on Ageing Management

Publication

2006 2006 --

③③③③Roadmaps

• Budgetary Actions forResearch and Infrastructure

• Regulation Systems withCodes and Standards

7

Major Roles of Industries, Government and Major Roles of Industries, Government and Academia for Ageing ManagementAcademia for Ageing Management

・・・・To develop, review strategy maps periodically and build a consensus with experts from industries, government and academia

by identifying important issues to be solvedby discussing roles of stakeholders for rational and efficient implementation of research

・・・・To develop and improve codes and standards based on safety research outcome

Academic Societies

• Responsibility to establish and improve safety regulations - by conducting safety researches, - by keeping infrastructure • Responsibility to promote nuclear energy for national energy security

Government

・Responsibility to contribute to safety research activities- to accumulate and expand fundamental knowledge- to keep human resources through the safety research

Academia

• Primary responsibility to keep safe operation of power plants

- research and development, and development of the infrastructure for safety and public interests

Industries

8

4. Systematic MaintenanceSystematic Maintenance3. Establishment of Information Basis3. Establishment of Information Basis

2. Establishment of Codes and Standards2. Establishment of Codes and Standards1. Safety Research to Solve Ageing Issues1. Safety Research to Solve Ageing Issues

* Advanced designs will be developed based on past operating experience.

* Predication methods and monitoring technology will be upgraded.

* Countermeasures are being taken by rule of thumb based on plant data.

Irradiation embrittlement

* Simulation methods will be established. * ISI technology will be upgraded.

* Use of SCC-resistant materials will be ensured. * Database will be constructed.

* Countermeasures suited to materials in use are being taken. * Database is being constructed.

Stress corrosion crack

* Past records of performance will be reflected in the maintenance technology and durability designs for new plants. * Recycling methods will be established for replaced structures and materials.

The reliability of integrity evaluation methods will be improved or enhanced. CCV integrity evaluation methods will be established.

Intensive study is being made on scarcely known fields.

Decreased strength of concrete

* Monitoring technology will be upgraded.

Deterioration diagnosis technology will be upgraded.

* Countermeasures suited to materials in use are being taken. * Database is being constructed.

Deteriorated cable insulation

* Monitoring technology will be upgraded.

* Mechanism-based predication methods will be established. * Risk-based maintenance methods will be established.

* Countermeasures suited to materials in use are being taken, and database is being constructed. Wall thinning

* Countermeasures suited to materials in use or environment will be taken, and database will be constructed.

* Countermeasures suited to materials in use or environment will be taken, and database will be constructed.

* Countermeasures suited to materials in use or environment are being taken, and database is being constructed. Fatigue

Long-life, next-generation reactor 40th to 50th year Current StatusAging

phenomenon

Strategy Maps 2007 for Ageing ManagementStrategy Maps 2007 for Ageing ManagementMaps for Technical IssuesMaps for Technical Issues

9

Strategy Maps 2007 for Ageing ManagementRoadmaps for Stress Corrosion CrackingRoadmaps for Stress Corrosion Cracking

10

Aps

Phase I (until 40 year operation of early stage NPPs2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015～～～～2019 2020～～～～2029

Phase II (until 50 year operation) Phase III (until 60 year operation)Fiscal Year

Preparation and refinement of initiation data and crack growth data on IASCC

Preparation of infrastructure for IASCC research Preparation and refinement of initiation data and crack growth data on IASCC Confirmation of the data and evaluation technology, and preparation of regulatory criteria

Understanding of IASCC mechanism

Simulation technology of IASCC initiation and propagation

Establishment of code and standards, and regulatory criteria

IASCC initiation test methods

Refurbishment of laboratory facility including JMTR

Extended IASCC crack initiation data and its refinement

Extended IASCC crack growth data and its refinementCorrelation between laboratory data and field events

Obtaining of crack growth data for ex-plant materials In-core IASCC Initiation and growth evaluation

IASCC evaluation technology confirmation Understanding and modeling of IASCC crack

Study of accelerated irradiation validity

Understanding and modeling of IASCC crack initiation and growth mechanism

Development of simulation technology of IASCC initiation and growth

Study of IASCC mechanism at high fluence region

Establishment of CS for fitness-for-service

IASCC integrity evaluation guide

Establishment of CS for fitness-for-serviceCoding of IASCC initiation and growth simulationSCC long term integrity evaluation guide

Establishment of code and standards, and regulatory criteria

1. Integrity Evaluation : IASCC1. Integrity Evaluation : IASCC2. Integrity Evaluation : IGSCC2. Integrity Evaluation : IGSCC

3. Integrity Evaluation : 3. Integrity Evaluation : NiSCCNiSCC4. Inspection Technologies on SCC4. Inspection Technologies on SCC

5. Monitoring Technologies on SCC5. Monitoring Technologies on SCC6. Repair & Replacement Technologies6. Repair & Replacement Technologies

Strategy Maps 2007 for Ageing ManagementRoadmap of Research and of Code Development for IASCCRoadmap of Research and of Code Development for IASCC

Phase I (until 40 year operation of early stage NPPs2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015～～～～2019 2020～～～～2029

Phase II (until 50 year operation) Phase III (until 60 year operation)Fiscal Year

Preparation and refinement of initiation data and crack growth data on IASCC

Preparation of infrastructure for IASCC research Preparation and refinement of initiation data and crack growth data on IASCC Confirmation of the data and evaluation technology, and preparation of regulatory criteria

Understanding of IASCC mechanism

Simulation technology of IASCC initiation and propagation

Establishment of code and standards, and regulatory criteria

IASCC initiation test methods

Refurbishment of laboratory facility including JMTR

Extended IASCC crack initiation data and its refinement

Extended IASCC crack growth data and its refinementCorrelation between laboratory data and field events

Obtaining of crack growth data for ex-plant materials

In-core IASCC Initiation and growth evaluation

IASCC evaluation technology confirmation Understanding and modeling of IASCC crack

Study of accelerated irradiation validity

Understanding and modeling of IASCC crack initiation and growth mechanism

Development of simulation technology of IASCC initiation and growth

Study of IASCC mechanism at high fluence region

Establishment of CS for fitness-for-serviceIASCC integrity evaluation guide

Establishment of CS for fitness-for-service

Coding of IASCC initiation and growth simulationSCC long term integrity evaluation guide

Establishment of code and standards, and regulatory criteria

11

10⑨

1714130908

T1

2F1 K4SK1

Tsu22F2０

O11F5

2

1F2⑦⑦⑦⑦

8

242322212019181615121110070605040302012000Year

O2K2TM11F42F41F3T4OG12F21F1Tsu1１F6１F4①①①①

1F3T21F2①①①①①①①①

Tsu1

I3K3M2GN2M1O1To2H3M32F3SD1M2②②②②②②②②

O2To2M3②②②②

T1②②②②

3

OG1④④④④④④④④④④④④

O2④④④④④④④④

SD21F2④④④④

5

SD1⑤⑤⑤⑤

GN1T16

9

GN3

2F2

SM1

O4

H4

I2

③③③③

TM2

1F1

K5

Tsu1

SM2

1F6

1F5

H2I1

H1

③③③③

T3

K1

T2

SM1

⑥⑥⑥⑥

I2

③③③③

GN2M1

③③③③

H2

GN1０

M21F1M1 I1H1SM11

③③③③③③③③

4

7

⑨⑨⑨⑨

12

Strategy Maps 2007 for Ageing ManagementScenarios(1)

-- Ageing Management Technical Evaluation for Every 10 Years Ageing Management Technical Evaluation for Every 10 Years --•• Increase in number of plantsIncrease in number of plants•• From initial plants to standardized plantsFrom initial plants to standardized plants

First plant to exceed 40 years of operation

First plant to exceed 50 years of operation

Note: This figure is drawn assuming that all of the plants currently in service will continue their operation.

99 00 01 02 03 04 05 06 07 08

2004-2006

1996-2003

Assessment of Cable Aging for NuclearPower Plant (ACA)

2003-2007

Evaluation Methodology of Crack GrowthRate for Ni-based Alloys (NISCC)Evaluation of Irradiation Assisted StressCorrosion Cracking (IASCC)

1997-2010

Nondestructive Inspection Technologies forthe Narrow Penetrations on Reactor Vessel(NPV)

2005-2008

2005-2009

Development of Seismic Safety EvaluationMethod on Aged Nuclear Facilities

2002-2008

2001-2007

Nondestructive Inspection Technologies onNi-based Alloy Welded Joint (NNW)

Nuclear Power Plant Material ImprovementTechnology (PMT)

Repair Welding Technology of IrradiatedMaterials (WIM)

Intergranular Stress Corrosion Cracking ofNuclear Grade Stainless Steel (IGSCC)

Nondestructive Inspection Technologies forCore Shroud Integrity Assessment (NSA)

2003-2006

2002-2008

Category

2000-2008

Priods

1996-20091994-2006

JFY

Assessment of concrete Aging for NuclearPower Plant

Nuclear Power Plant Integrated ManagementTechnology (PLIM)

2000-2009

Environmental Fatigue Tests of NPPMaterials for Reliability Verification (EFT)Integrity Assessment of Flawed Componentswith Structural Discontinuity (IAF)

Major Projects

Agein

g Eva

luatio

nIns

pecti

on

and

Monit

oring

Evalu

ation

of SC

C

Maint

e-na

nce&

Re

pair

JNES Projects for Ageing Management JNES Projects for Ageing Management Sa

fety R

esea

rch on

evalu

ation

of N

PPSe

ismic

safet

y

13

100 1000 10000 1000000

100

200

300

400

500

80oC-3Gy/h 90oC-3Gy/h 100oC-3Gy/h 80oC-18Gy/h 90oC-18Gy/h 100oC-18Gy/h 80oC-100Gy/h 90oC-100Gy/h 100oC-100Gy/h

Elong

ation

at br

eak (%

)

Time (hours)

Simultaneous aging characteristics of the FR-EPR insulator (black core) made by A company

100 1000 10000 1000000

100

200

300

400

500

600

80oC-3Gy/h 90oC-3Gy/h 100oC-3Gy/h 80oC-18Gy/h 90oC-18Gy/h 100oC-18Gy/h 80oC-100Gy/h 90oC-100Gy/h 100oC-100Gy/h

Elong

ation

at br

eak (%

)

Time (hours)

Simultaneous aging characteristics of the XLPE insulator made by A company

Examples of the results obtained from ACA projectExamples of the results obtained from ACA project

� Progress of degradation may be significantly different among various manufacturers, even for the same kind of insulator. � There were some cables that indicated significantly rapid progress of

degradation at low dose during simultaneous aging, and many other cables also indicated relatively rapid progress of degradation during simultaneous aging.

Major results obtained from ACA project

14

Elong

ation

at Br

eak (

%)

Elong

ation

at Br

eak (

%)

Time (hrs) Time (hrs)

PLM research promotion council board 【【【【Policies regarding research promotion】】】】

［［［［Secretariat: JANTI (FEPCO)］］］］

PLM research administrative review meeting

【【【【Discussing research strategies】】】】［［［［Secretariat: CRIEPI］］］］

10 Sub Group (SG): Examination team【【【【Individual Issues.】】】】

Providing future directionsEvaluation results

Industrial Efforts to Promote PLMIndustrial Efforts to Promote PLM--related R&D Activities *related R&D Activities *

＜＜＜＜Electric power companies, manufacturers and organizations that are involved in the support system＞＞＞＞

Nine electric power companies, JAPC, Electric Power Development CO., CRIEPI and JANTI

Technology information coordination committee

Collaboration

Collaboration

Nuclear energy development committee Comprehensive

task force

Nuclear energy development committee

Equipment maintenance committee

＜＜＜＜The Federation of Electric Power Companies of Japan＞＞＞＞

Chairman and deputy chairman of promotion council board participate in the committee.

15

* * KamadaKamada et al; this Symposiumet al; this Symposium

Research ProjectResearch Project on Fundamentals ofon Fundamentals ofAgeing ManagementAgeing Management (NISA funds)(NISA funds)

General Review Session on Ageing Management Infrastructure Improvement

Review Session on Ageing Management Database and Knowledge Base Review Session on Seismic Effects on Piping ThinningReview Session on Condition Monitoring and Safety Preservation Technology Review Session on Irradiation Embrittlement

Review Session on Cable Degradation

Review Session on Stress Corrosion Cracking

Tohoku ClusterICR

Fukui ClusterINSS

East Japan Cluster

MRI

Ibaraki ClusterJAEA

General Review Session WG

InformationExchange

ProgressReportAdvice

ProgressReportAdvice

TechnicalInformationCoordinationCommitteeSafety ResearchWGIn JNES

Outcome

NISA (Nuclear and Industrial Safety Agency) in METI

Academic Societies etc.

InformationExchange

16

Development of New Embrittlement CorrelationDevelopment of New Embrittlement Correlation

• Fundamental study on the embrittlement mechanisms– Nano-structural characterization by experiments– Multiscale computer simulation of microstructural evolution

• Mechanism-guided embrittlement correlation– Surveillance data in Japan– New approach using rate equations in predicting microstructural changes

17

Hierarchical Structure of Degradation ofHierarchical Structure of Degradation ofMaterials for Light Water ReactorsMaterials for Light Water Reactors

Effects of Radiation in Nuclear Materials and Effects of Radiation in Nuclear Materials and Ageing Management of Nuclear SystemsAgeing Management of Nuclear Systems

Radiation Damage ProcessesRadiation Damage Processes

Evaluation MethodologiesEvaluation Methodologies

Nuclear Reaction

Phase10 sec

-13

10 sec

-11

10 sec

-6

M onte-C arlo

Sim ulation

10 sec

9

10 sec

-3

10 sec 10 sec

14～7

Collisional Phase

Cooling Phase

Thermal Phase

Diffusional Phase

Micro- structure

Property Changes

Lifetime Optimization

Binary Collision Approximation

Molecular Dynamics

Analytical Diffusion & Reaction Rate Equations

Dislocation Dynamics

Finite Element Method

Post Irradiation Tests

System Maintenance, Safety, Reliability Economy

Single CascadeMultiple Cascades, Subcascades Cascade Clusters,

Migrating DefectsNucleation and Growth of Defect Clusters

Volume Change, Strength, Creep, Fatigue, Corrosion, Crack Initiation & Growth ....

10 sec

-15

NPRIM Code

Primary Knock-on Atom

Nuclear Reaction

Phase

N. Sekimura (1994)18

MultiMulti--scale Computer Simulationscale Computer SimulationMolecular DynamicsDislplacement cascade

Kinetic Monte CarloMicrostructural evolution during irradiation

Dislocation DynamicsDislocation behavior during deformation

Detailed analysis of microstructure

Point defect production

Cu atoms

Vacancies

Dislocationloop

Dislocation

Radiation damage

Molecular Dynamics

Molecular Dynamics

Stres

s (MP

a)

Strain (%)

~10-11sec~10-8m

~109sec~10-7m

~100sec~10-4m

Dislocation DynamicsPrediction of mechanical property ~100m

Unirradiated

Irradiated

Interaction between dislocation and damage

19

NanoNano--structural Characterizationstructural Characterization

3-Dimensional Atom ProbePositron Annihilation

(Coincidence Doppler Broadening)

1.4

1.2

1.0

0.8

0.6

Rat

io to

Fe

50403020100PL (x10-3 m0c)

Unirrad. As-irrad. 500°C

規格

化され

た強度

電子の運動量

未照射材照射材

熱時効材

Cu-enriched clusters formed by neutron irradiation

~40 nm

~300

nmLEAP

(Local Electrode Atom Probe)

Norm

alize

d cou

nts of

gamm

a ray

s

Electron momentum

Irradiated unirradiated

thermallyaged

20

New Embrittlement Correlation for JEAC 4201New Embrittlement Correlation for JEAC 4201

MDMD CT ⋅=∆ 18ξ

( ) ( )22MDSC TTT ∆+∆=∆

( ) ( ) ( ) SCNiSCmatCufSC CthCgCCfVT ⋅+⋅⋅⋅=⋅=∆ φξξξ 0

161717 ,

( )( ) ( )( )2089214 1 NiCu

availCuMDCu

matCu

SC CDCCDCt

C ⋅+⋅⋅⋅+⋅+⋅+⋅=∂

∂ ξξξξξ

( )t

CCFt

C SCNit

MD

∂∂−⋅⋅+⋅⋅=

∂∂ φξξξ 2

762

5

SCSC

enhSc

SC

matCu Cv

tCv

tC ⋅′−

∂∂⋅−=

∂∂

21

ηφη ⋅+=+= thermalCu

irradCu

thermalCuCu DDDD

( )t

CCFt

C SCNit

MD

∂∂−⋅⋅+⋅⋅=

∂∂ φξξξ 2

762

5

21

プラント補正なし

プラント補正あり

-20

0

20

40

60

80

100

120

140

-20 0 20 40 60 80 100 120 140

監視試験測定値（℃）

予測値

（℃）

補正なし

補正あり

1:1

-2σ

+2σ

0.15.4CRIEPI adj0.79.4CRIEPI2.311.7E900-022.810.4EWO-1.915.4RG1.99ｒ2-1.311.9JEAC4201

Mean ErrorStd. Dev.Method

Pred

iction

(oC)

Measured value (oC)

w/o adjustmentw adjustment

∆T

φtOffset

∆T

φtOffset

Plant specific adjustment

Comparison of DBTT shift betweenComparison of DBTT shift betweenthe measured value and the predictionthe measured value and the prediction

22

RVH nozzle environmental shielding weld (seal welding over cracking)

RVH nozzle penetration repair technique

Overlay weldOverlay weld

The structure and application of Inconel 690 repair weld have been approved by the verification committee, RTTW committee and CRT

committee.

Cracking

Applying corrosion resistant Inconel 690 overlay weld directly over through-wall cracking

Repair welding machine

Repair weld applied to through-wall cracking in J-groove weld

Through-wall cracking

Repair Welding Technique Development; MHIRepair Welding Technique Development; MHI

weld

Weld device

三菱重工原子力技術資料：クラスB

All Rights Reserved, Copyright (C), MITSUBISHI HEAVY INDUSTRIES,LTD. 23

History of Troubles and Preventive Maintenance of RVHHistory of Troubles and Preventive Maintenance of RVH

▼

１９８０ １９９０ ２０００

Trouble

R/V＜head penetration nozzle＞＜BMI nozzle＞

Present

S/G＜Tube＞

SCC of tubes of S/G in Japan (MA600)

SCC of tubes of S/G in Japan (TT600)

SCC of nozzle of pressurizer at Tsuruga unit-2 (2003)

SCC of R/V head nozzle at Bugey unit-3 in France(1991) ▼

SCC of R/V outlet nozzle at V.C.Summer in the U.S. (2000) ▼

SCC of R/V head nozzle at Oconee unit-1 in the U.S. (2000) ▼ SCC of BMI nozzle at South Texas Project unit-1 in the U.S. ▼ (2003)

ECT of R/V head penetration nozzle

Replacement of R/V head

R/V,Pressurizer ,S/G＜Safe-end welds of nozzle＞

Preventive Maintenance

Reduction of R/V head top temperature(Takahama3.4,Ohi3.4)

Intelligent ECT of tube of S/G (TT600)Replacement of S/G(MA600→TT690）

Shot-peening of tubes of S/G (TT600)

WJP / ECT of BMI nozzle (base metal)

WJP of R/V inlet/outlet nozzleShot-peening of S/G inlet/outlet nozzleLaser stress improvement of Pzr nozzle

WJP of BMI nozzle(weld metal)

UT (angle beam,longitudinal wave) of safe-end welds of nozzle (R/V,Pressurizer,S/G)

SCC of nozzle of R/V head nozzleat Ohi-3 (2004) ▼

SCC of BMI nozzle at Takahama-1(2002) ▼

Replacement of R/V head(Takahama3.4,Ohi3.4)

24

Reactor Vessel Head Nozzle MaintenanceReactor Vessel Head Nozzle Maintenance200520001995 2010 2015

ECT

▼Bugey 3Leak(1991)(France）

VHR

T-Cold VHR

▼Davis-Besse Leak(2002)

▼Ohi3 Leak(2004)

Older PlantsMihama 1,2,3Takahama 1,2Ohi 1,2

New PlantsTakahama 3,4Ohi 3,4

Preventive maintenance (T-Cold) has been conducted at new plants with a lower possibility of PWSCC in order to mitigate PWSCC susceptibility.

ECT had been conducted at old plants with a relatively high possibility of PWSCC.

VH had been replace at old plants with a relatively high possibility of PWSCC.

Replacement of VH is also planned for new plants.

PlugPlug hole Increased plug hole diameter

Change of nozzle Base & J-Weld material. Alloy 600 Alloy 690(Improved corrosion resistance)

25

1. Purpose・Mitigation of SCC for pipes orcylindrical equipments

2. Effect・Residual Stress Improvementl for weld joints and HAZ onthe inner surface

3. Experiences・BWR Plants in Japan

HAZ : Heat Affected Zone

Induction Heating Stress Improvement（（（（IHSI））））

PipePipe

CoolantCoolantCoolantCoolant

IH coil

All Rights Reserved Copyright © 2007, Hitachi-GE Nuclear Energy, Ltd.

26

Controller

Laser systemOptical fiber

CRD stub tube

Remote handlingsystem

Shroud

Preventive Maintenance Technology DevelopmentPreventive Maintenance Technology Development

�Development of a laser peeningsystem for core shroud.�Application for Japanese nuclear reactors (BWR) since 1999.�Development and application of a fiber-delivered system since 2002.�Experiences in BWRs: 8 plants.

CRD Stub Tubes> 4 plantsCore Shroud>5 Plants

�PWR Applications since 2004:2 plants.

Copyright © 2006 Toshiba Corporation. All rights reserved.

Fiber-delivered Laser Peening SystemApplication for CRD Stub Tubes （（（（BWR））））

27

Recent activity and LP System architecture Recent activity and LP System architecture

Road Map: Examples of R&D and Good Practices Road Map: Examples of R&D and Good Practices

28

1F-1～～～～5: Early

SCC of type 304 piping, oversea & domestic NPP

SCC of type 316L▼▼▼▼

at PLR Piping, H-3 NPP

▼▼▼▼

▼▼▼▼UT indicationnozzle, oversea NPP

SCC of type 304Lshroud, oversea NPP

SCC of type 316L piping, domestic NPP

‘70 ‘80 ‘90 ‘00Major Events

Materials Improvements

Stress Improvements

Environment Improvements

Replacement Repair

Coun

term

easu

res

Major R&D

Type 304 NP 1F-6, 2F-1After NP 2F-2

Type 304LType 316(LC)

Induction Heating Stress Improvement, Heat Sink Welding, Corrosion Resistance CladdingNP 1F-6, 2F-1 Replace: 1F-1～～～～5

All plants

Start-up deaeration Operation

Hydrogen injectionNoble metal chemical addition

1F-1～～～～6、、、、2F-3 etc.

2F-3

Except main piping Main piping To expand whole type 316L

1F-1～～～～5 (Type 304+Stress improve)Development of low carbon SS

Development of Countermeasures to SCC

Development of Hydrogen injection

1F-1～～～～5

PMT, WIM(JAPEIC/JNES projects)

Experiences and Countermeasures of SCC in PLR PipingExperiences and Countermeasures of SCC in PLR Piping

Proprietary Information

Cooperation with International OrganizationsCooperation with International Organizations• IAEA

Ageing Management Guideline (for aged NPP)• OECD/NEA

SCAP SCC and Cable Ageing ProjectIAGE Metal-Concrete-Earthquake Resistance-HF Steering GroupOPDE OECD Pipe failure Data Exchange Project

• USANRC PINC ProjectASME Sec.XI Code CommitteeIEEE Cable Environmental Qualification Test

• IECRadiation Effect Interpretation (Judgment) Criteria

29

Davis-Besse Reactor Vessel Head Degradation

Project Organization

Steering Committee

U.S. Nuclear Regulatory Commission (USNRC) proposed international cooperative research Project for the ＩＩＩＩnspection of Nickel-alloy Components (PINC). And Five country eight organizations

are participating in PINC to address common problem.

TG-AtlasDevelop an electronic Atlas (database) of NDE and metallography information TG-NDERound Robin of NDE techniques on PWSCC and simulated cracks

PINC Project

(2004~2008)

30

Round Robin Test in TG-NDE

Performance to detect PWSCC will be clarified and the results will be reflected to regulatory guidelines andprocess qualifications by NDI test of CRDM, BMI, Safe-end dissimilar metal welds.

Sample of CRDM TPs Sample of BMI TPs Sample of DMW TPs

PINC Project

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OECD/NEA OECD/NEA SCAP ProjectSCAP Project **

• The main objectives of SCAP are to:– Establish a complete database with regard to major ageing phenomena for SCC and degradation of cable insulation through collective efforts by OECD/NEA members– Establish a knowledge-base by compiling and evaluating collected data and information systematically, with regard to major ageing phenomena for SCC and degradation of cable insulation– Perform an assessment of the data and identify the basis for commendable practices which would help regulators and operators to enhance ageing management

SCAP: Stress Corrosion Cracking and Cable Ageing Project

*This project is being financed through Japanese voluntary contribution 32

(Members)・Belgium・Canada・Czech Republic・Finland・France・Germany・Japan・Korea・Mexico・Norway・Slovak Republic・Spain・Sweden・U.S

Management Board

Cable WGSCC WG

(Observers) IAEA, EC

MB Chair Prof.Sekimura

(Japan)

SCC Chair Mrs. Gott (Sweden)

SCAP Consultant Mr. Shultz (Germany)

Cable Chair Mr. Koshy (U.S)

Clearing House

ERIN（U.S)

Clearing HouseOECD Halden

Reactor ProjectIFE(Norway)

Management of the SCAPManagement of the SCAP

Technical Secretariat 33