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NATIONAL ASSESSMENT REPORT OF FINLAND

for the Purposes of Topical Peer‐Review “Ageing Management”

under the Nuclear Safety Directive 2014/87/EURATOM

RadiationandNuclearSafetyAuthority Report NuclearReactorRegulation December29,2017

Contents

1 Generalinformation............................................................................................................................................................3

1.1 Nuclearinstallationsidentification.....................................................................................................................3

1.2 Processtodevelopthenationalassessmentreport.....................................................................................3

2 Overallageingmanagementprogrammerequirementsandimplementation...........................................4

2.1 Nationalregulatoryframework............................................................................................................................4

2.2 Internationalstandards............................................................................................................................................4

2.3 Descriptionoftheoverallageingmanagementprogramme....................................................................5

2.3.1 ScopeoftheoverallAMP................................................................................................................................5

2.3.2 Ageingassessment............................................................................................................................................7

2.3.3 Monitoring,testing,samplingandinspectionactivities...................................................................9

2.3.4 Preventiveandremedialactions.............................................................................................................13

2.4 ReviewandupdateoftheoverallAMP...........................................................................................................14

2.5 Licensee’sexperienceofapplicationoftheoverallAMP.........................................................................14

2.6 Regulatoryoversightprocess.............................................................................................................................15

2.7 Regulator’sassessmentoftheoverallageingmanagementprogrammeandconclusions.......15

3 Electricalcables..................................................................................................................................................................17

3.1 Descriptionofageingmanagementprogrammesforelectricalcables.............................................17

3.1.1 Scopeofageingmanagementforelectricalcables...........................................................................17

3.1.2 Ageingassessmentofelectricalcables..................................................................................................20

3.1.3 Monitoring,testing,samplingandinspectionactivitiesforelectricalcables.......................22

3.1.4 Preventiveandremedialactionsforelectricalcables....................................................................26

3.2 Licensee’sexperienceoftheapplicationofAMPsforelectricalcables.............................................27

3.3 Regulator’sassessmentandconclusionsonageingmanagementofelectricalcables...............29

4 Concealedpipework.........................................................................................................................................................30

5 Reactorpressurevessels................................................................................................................................................31

5.1 DescriptionofageingmanagementprogrammesforRPVs...................................................................31

5.1.1 ScopeofageingmanagementforRPVs.................................................................................................32

5.1.2 AgeingassessmentofRPVs........................................................................................................................36

5.1.3 Monitoring,testing,samplingandinspectionactivitiesforRPVs..............................................46

5.1.4 PreventiveandremedialactionsforRPVs..........................................................................................55

5.2 Licensee’sexperienceoftheapplicationofAMPsforRPVs...................................................................58

5.3 Regulator’sassessmentandconclusionsonageingmanagementofRPVs.....................................62

6 Calandria/pressuretubes(CANDU)..........................................................................................................................64

7 Concretecontainmentstructures...............................................................................................................................64

RadiationandNuclearSafetyAuthority Report NuclearReactorRegulation December29,2017

7.1 Descriptionofageingmanagementprogrammesforconcretestructures......................................64

7.1.1 Scopeofageingmanagementforconcretestructures....................................................................64

7.1.2 Ageingassessmentofconcretestructures...........................................................................................71

7.1.3 Monitoring,testing,samplingandinspectionactivitiesforconcretestructures................76

7.1.4 Preventiveandremedialactionsforconcretestructures.............................................................87

7.2 Licensee’sexperienceoftheapplicationofAMPsforconcretestructures......................................90

7.3 Regulator’sassessmentandconclusionsonageingmanagementofconcretestructures........96

8 Pre‐stressedconcretepressurevessels(AGR).....................................................................................................98

9 Overallassessmentandgeneralconclusions.........................................................................................................99

9.1 Overallageingmanagement................................................................................................................................99

9.2 Electricalcables........................................................................................................................................................99

9.3 Reactorpressurevessels...................................................................................................................................100

9.4 Concretecontainmentstructures...................................................................................................................101

10 References....................................................................................................................................................................103

11 ANNEX1:FinnishNPPsites..................................................................................................................................105

12 ANNEX2:KeytechnicaldocumentationofNPPunitsOlkiluoto1and2cabling...........................106

13 ANNEX3:KeytechnicalOL3projectdocumentationrelevanttocabling..........................................107

14 ANNEX4:ASchematicdrawingsofLoviisa1andLoviisa2RPVs........................................................108

15 ANNEX5:AschematicdrawingofOlkiluoto1andOlkiluoto2RPVs..................................................109

16 ANNEX6:AschematicdrawingofOlkiluoto3RPV...................................................................................110

17 ANNEX7SummaryofOlkiluoto3RPVsubcomponents’materials.....................................................111

18 ANNEX8:AschematicdrawingofLoviisa1and2reactorbuilding...................................................113

19 ANNEX9:AschematicdrawingofOlkiluoto1and2containment.....................................................114

20 ANNEX10:AschematicdrawingofOlkiluoto3reactorbuilding........................................................115

RadiationandNuclearSafetyAuthority Report 1 (115) NuclearReactorRegulation December29,2017

Abbreviations

312 Feedwatersystem321 Shutdowncoolingsystem323 ReactorcorespraysystemAMP AgeingmanagementprogrammeAPC AirplanecrashBWR BoilingWaterReactorCASS CastausteniticstainlesssteelsCFD ComputationalFluidDynamicsEPW ExplosionpressurewaveFAC Flow‐acceleratedcorrosionggbs GroundgranulatedblastfurnaceslagOPC OrdinaryPortlandcementPH Precipitation‐hardenedCRDM ControlroddrivemechanismCUF CumulativeusagefactorDMW DissimilarmetalweldECCS EmergencycorecoolingsystemECP ElectrochemicalPotentialENSREG EuropeanNuclearSafetyRegulatorsGroupIAEA InternationalAtomicEnergyAgencyIASCC IrradiationacceleratedStressCorrosionCrackingIGSCC IntergranularStressCorrosionCrackingIGALL IAEA’sInternationalGenericAgeingLessonsLearnedIRWST In‐containmentRefuellingWaterStorageTankKTO PeriodicinspectionprogrammeLO1 NPPunitLoviisa1LO2 NPPunitLoviisa2LOCA LossofcoolantaccidentLTO LongtimeoperationMCL MaincoolantlineMWe MegaWattelectricpowerMWth MegaWattthermalpowerNAR NationalAssessmentReportNPP NuclearPowerPlantNSD NuclearSafetyDirectiveOL1 NPPunitOlkiluoto1OL2 NPPunitOlkiluoto2OL3 NPPunitOlkiluoto3PAMS PipingandcomponentAnalysisandMonitoringSystemofTVOPSR PeriodicSafetyReviewPWR PressurizedWaterReactorR&D ResearchandDevelopmentRCPB ReactorcoolantpressureboundaryRCSL ReactorControl,SurveillanceandLimitationRI‐ISI Riskinformedin‐serviceinspection


RL WENRAReferenceLevelforExistingReactorsRPV ReactorpressurevesselSAFIR TheFinnishResearchProgrammeonNuclearPowerPlantSafetySC SafetyclassSCC StresscorrosioncrackingSS StainlesssteelSSC System,StructureorComponentimportanttosafetySSE Safeshut‐downearthquakeSTUK RadiationandNuclearSafetyAuthorityTH LowpressuresafetyinjectionsysteminLoviisaTPR TopicalPeerReviewUT UltrasonictestingVVER AtypeofPWR(WaterWaterEnergeticReactor)VT VisualtestingVTT TechnicalResearchCentreofFinlandWANO WorldAssociationofNuclearOperatorsWENRA WesternEuropeanNuclearRegulatorsAssociationYTN STUK’sAdvisoryCommitteeonNuclearSafety


1 Generalinformation

1.1 Nuclearinstallationsidentification

TheLoviisaNuclearPowerPlant(NPP)housestwoSoviet‐designed(Atomenergoexport)VVER‐440/213pressurizedwaterreactors(PWRs),i.e.,Loviisaunit1andLoviisaunit2.Thecurrentcapacityaftersomemodernisationsis2x502MWe.AgeneralviewofLov‐iisaNPPispresentedinAnnex1.TheLoviisaNPPunitsstartedcommercialoperationin1977(Loviisa1)and1981(Loviisa2)respectively.TheplantisoperatedbyFortumOyj.

TheOlkiluotoNPPconsistsoftwoBoilingWaterReactors(BWRs),i.e.,Olkiluotounit1andOlkiluotounit2andonePWR,i.e.,Olkiluotounit3.ThecurrentcapacityofOlkiluotounits1and2isafterseveralmodernisations2x880MWe.TheplannedcapacityofOlki‐luotounit3is1600MWe.AgeneralviewofOlkiluotoNPPispresentedinAnnex1.Olki‐luoto1startedcommercialoperationsinOctober1979andOlkiluoto2inJuly1982.Thedesigneroftheunits1and2wasSwedishAsea‐AtomwhichnowadaysbelongstoWest‐inghouse.

Olkiluotounit3isEPR‐type(EuropeanPressurizedReactor),designedbyaconsortiumofArevaandSiemens.TheconstructionlicensewasgrantedbytheGovernmentinFeb‐ruary 2005. The plant unit is currently in commissioning phase. Current schedule issuchthatthefuellingisexpectedinAugust2018andthestartofcommercialoperationinMay2019.

OlkiluotoNPPisownedandoperatedbyTeollisuudenVoimaOyj(TVO),asubsidiaryofPohjolanVoimaOyj.

TheresearchreactorFIR‐1oftypeTRIGAMarkIIislocatedinOtaniemiCampusarea.Itwaspurchased for researchpurposes and itwas started in1962. Later on itwas alsoused forproducing isotopes for industryandmedicalpurposes. Its thermalcapacity is250kW.TheresponsibleorganizationwasoriginallyHelsinkiUniversityofTechnologyandlateron(after1971)VTTTechnicalResearchCentreofFinlandLtd.Thisreactorisnot inthescopeof thisNARdueto its lowthermalpower.Furthermore, thereactor isnowpermanentlyshutdownandpreparingfordecommissioning.

1.2 Processtodevelopthenationalassessmentreport

CoordinationofNARpreparationinFinlandwasdonebytheRadiationandNuclearSafe‐tyAuthority (STUK).ForNARpreparationacross‐sectionalworkinggroupwassetuprepresenting three different disciplines, i.e., I&C, electrical, mechanical and civil engi‐neering,andhavingknowledgeandexperiencealsointheareaofagingmanagement.

ThelicenseeswereinvitedtosupplymaterialsforthepreparationofNAR.TheNARwaspreparedon thebasisof thesedocumentsand thecontributions fromthemembersoftheworkinggroup.MostofthematerialwasalreadyearlierdeliveredtoSTUKbutalsosomenewmaterialwasdeliveredbythelicenseesduringtheNARpreparationprocess.

TheNARwaswrittendirectlyinEnglish.ThedraftofNARwassubjectedtocommentingprocedureandcommentswereaskedfromthelicenseesandSTUK’sstaffmembers.The


draftoftheNARwasalsopresentedtotheSTUK’sAdvisoryCommitteeonNuclearSafety(YTN),whosemembershadopportunitytogivetheircomments.

TheNARistobedeliveredtotheENSREGandpublishedattheSTUK’swebsiteafteritsfinalization.

2 Overallageingmanagementprogrammerequirementsandimplementation

2.1 Nationalregulatoryframework

STUK Regulation on the Safety of a Nuclear Power Plant (STUK Y/1/2016) stipulatesthatthedesign,construction,operation,conditionmonitoringandmaintenanceofanu‐clear power plant shall provide for the ageing of systems, structures and components(SSCs) important to safety inorder toensure that theymeet thedesign‐basis require‐mentswithnecessarysafetymarginsthroughouttheservicelifeofthefacility.Further‐morereferringtotheSTUKRegulation,systematicproceduresshallbeinplaceforpre‐ventingsuchageingofSSCswhichmaydeterioratetheiroperability,andfortheearlyde‐tectionoftheneedfortheirrepair,modificationandreplacement.Safetyrequirementsandapplicabilityofnewtechnologyshallbeperiodicallyassessedinordertoensurethatthetechnologyappliedisuptodate,andtheavailabilityofthesparepartsandthesys‐temsupportshallbemonitored.

Anewregulatoryguide[GuideYVLA.8AgeingManagementofaNuclearFacility]bothimposes requirements on licensees related to themanagement of physical ageing andobsolescenceofSSCs,andpresentstheregulatoryoversightrelevanttothelicensees’du‐ties.TheGuideappliestoallNPPlifecyclephasesandallSSCsimportanttonuclearandradiation safety. Regulatory requirements set in theGuide aim at ensuring both shortandlongtermoperabilityandtechnologicalconformanceofSSCswhetherinserviceorstand‐by. The key documents for regulator’s review are a conceptual plan for ageingmanagementandanageingmanagementprogrammealongwithaconstructionandop‐eratinglicenseapplications,respectively,andageingmanagementfollow‐upreportsis‐suedannuallybythelicensees.

2.2 Internationalstandards

ThebasicrulesofinternationalstandardsandguidelinesarefollowedinregulatinganddevelopingageingmanagementofFinnishNPPs,including:

WENRASafetyReferenceLevels forExistingReactors; Issue I:AgeingManage‐ment;

SafetyofNuclearPowerPlants:Design;IAEANoSSR‐2/1;

SafetyofNuclearPowerPlants:CommissioningandOperation;IAEANoSSR‐2/2;

AgeingManagementforNuclearPowerPlants;IAEANoNS‐G‐2.12.

The mentioned guidelines are adapted to Finnish maintenance strategies and ageingmanagementpracticestoanextentwhichhasbeendeemedappropriatetoensureoper‐abilityofSSCsatNPPs.Theirutilizationinthedevelopingprocessofageingmanagementisaddressedinthefollowingparagraphs.


2.3 Descriptionoftheoverallageingmanagementprogramme

2.3.1 ScopeoftheoverallAMP

TVO

Ageingmanagementiscoordinatedbyadedicated“Age”workinggroupwhichconsistsofrepresentativesofall technicaldisciplinesandrelevantplants’ functionssuchasnu‐clear safety, operation, maintenance and asset management. The main duties of thisgroupinclude

processinginformationrelatedtotheconditionsandperformanceofSSCs;

keepingupalifecycledatabaseforSSCs(recommendedmajormodifications,re‐placements,repairsandoverhaulswithinnext20years);

coordinatinginputinformationforinvestmentandoutageplanning;

revisingdocumentationforplantageingmanagement.

Apersonincharge“systemresponsible”isappointedforeachplantsystem.Thispersonisfamiliarizedwithbothoperationandsafetyfunctionsofhissysteminallplantopera‐tionmodes.Hismaindutyistoanalysetheperformanceandsafetymarginsofthesys‐tem,and to commithimselfon theneeds fordevelopment related tosystemmodifica‐tionsorthescopeandfrequencyofinspectionsandtests.

SSCs are divided into groups and a person in charge “component responsible” is ap‐pointedforeachgroup.Thispersonisakeyexpertwhoisfamiliarwithfunction,opera‐bility requirements and maintenance of plant components he is responsible for. Hekeeps record of maintenance works, inspections and tests, failure trends, operationhoursetc.Furthermore,thecomponentresponsiblemonitorsoperabilityofhiscompo‐nent group and takes appropriate actions whenever operability is endangered in theshortor longperiod.Hecollectsall information relevant to ageingmanagementofhiscomponentgroupanddrawsupastatusreportonaregularbasis.

Boththesystemandthecomponentresponsiblepersonscommunicatewithandprovideessential information to theAGEgroup toensure that theoverallAMP isbeing imple‐mentedandcontinuouslydeveloped.

Identifying SSCswithin the scope of overall AMP follows the SSC safety classification.Mechanical, electrical, I&Cor civil system, structure or components arewithin overallAMPiftheyareclassifiedtosafetyclass1,2or3.Inaddition,somenon‐nuclearsafetyclassifiedSSCshavinganconsequentialrisktonuclearsafetyareincludedintheoverallAMP inFinland.SuchSSCsmay initiateaneventby their failureor theyprotectsafetyfunctions against internal or external threats. In order to assure the integrity or func‐tionalcapabilityofaparticularSSCwithintheoverallAMP,themeasuresforthisassur‐ancedependontheSSC’ssignificancetonuclearsafetyandavailability forpowerpro‐duction.ForthispurposeSSCsareassignedtofourmaintenancecategories:category1“keepalwaysoperable”,category2“limitedunavailabilityallowed”,category3“econom‐icallyjustifiedpreventivemaintenanceallowed”andcategory4“nopreventivemainte‐nance”.Probabilisticriskanalyses,includingsuchimportantmeasuresasFussell‐Vesely


andBirnbaum,theOperationalLimitsandConditions,gatheredmaintenanceexperienceetc.havebeenutilizedwhencategorizingtheSSCsandplanningappropriateactionstomaintaintheiroperability.

TheAGEworkinggrouphasthekeyroleinthequalityassuranceoftheoverallAMP.In‐ternalauditsaccordingtotheQualityManagementSystemofthelicenseearemade,andifnecessary,externalevaluationsandregulatoryinspectionswillbearranged.TheAGEworking groupmake regular overviewsof the inspection and test results andmainte‐nance data, assesses the effectiveness of the process and then decides on remedialmeasuresiffailuretrendsarerisingoracceptancecriteriaaregettingclosetotheirlim‐its.Risinglong‐termfailuretrendsaregenerallycountedasthemostunambiguousindi‐catorthatrevealspossibleweaknessesintheimplementationoftheoverallAMP.

Fortum

The lifetime management group of the engineering and maintenance division is inchargeofageingmanagementatLoviisaNPP.OntheSSClevel,systemengineersareap‐pointedforeachplantsystem,andwithinthededicatedplantsystems,theirdutiesare

identificationandfollow‐upofdegradationmechanism;

managementofobsolescence;

maintainingofdatasystems(LOAM,POMS)forageingmanagement;

organizingofconditioninspectionsandmodificationworks;

preparationofinvestmentplans;

managementofSSCqualifications;

contactstointernalandexternalinterestgroups;

maintainingofSSCconditionclassification;

inputtoprocurementofsparepartsandsparepartstockstrategy;

plantwalkdowns.

Ageingmanagementcoversallsafetyclassifiedmechanical,electrical,I&CandcivilSSCs(system,structureorcomponent)withinoverallAMPiftheyareclassifiedtosafetyclass1,2or3.Alsonon‐nuclearsafetyclassifiedSSCsareincludedintheoverallAMPiftheyareconsideredtocauseanconsequentialrisktonuclearsafety.SSCsareclassifiedintothree groupsA, B andC having gradedprocedures and scopes of ageingmanagementeach.TheSSCsthatareassumedto limittheplant lifetimearecountedtogroupA,theSSCsthatarehighlysignificanttoavailabilityorsafetyoftheplantarecountedtogroupBandrestoftheSSCswithintheoverallAMPbelongtogroupC.

Furthermore,alltheNPPcomponentsareclassifiedintofourcriticalityclassesbasedontheir significanceonnuclearsafetyandpowerproduction.Thecriticalityclassesare1


(highcritical),2(critical),3(non‐critical)and4(runto failure).Theclassificationwasoriginallyintroducedtooptimizemaintenancetasksbutnowutilizedintheageingman‐agementofSSCs,too.

2.3.2 Ageingassessment

TVO

IAEANoNS‐G‐2.12andtheFinnishregulatoryguideYVLA.8havebeenthemainguid‐anceastheoverallAMPhasbeenprepared.

Plant programmes, i.e., scheduled maintenance and follow‐up programmes form thetechnicalbasisofageingassessment.Theplantprogrammesaretheleadingproceduresfortheidentificationofageingmechanismsandtheirpossibleconsequences.Theevalua‐tionof theirperformanceforexample intermsofSSCfailuretrends indicate ifappliedprogrammesareadequateeitherassuchormodifiedtomanagetheadverseageingef‐fectsforSSCs.

Scheduledmaintenanceprogrammesare implementedtoSSCsbelongingmainly tothemaintenancecategories1 to3.Themaintenancetaskswithin thecategoriesaredeter‐minedbySSC’simportancetonuclearsafetyandpowerproduction.Themostcompre‐hensive scheduledmaintenance in the category 1 is typically executed to SSCswithinOperationLimitsandConditionsoftheNPP.Duringthescheduledmaintenance,orun‐scheduled, too, conditions of a SSC is monitored and feedback from themaintenanceworksisprocessedforanyimprovements.Forexample,replacementperiodsforsparepartsandconsumablesorperiodsbetweenthescheduledinspectionsandtestsmaybereconsideredbasedonthefeedback.

Variousfollow‐upprogrammesareexecutedintheassessment,including

loadmonitoring(mechanicalcomponents);

loadingandstresscalculations(pipingandsupports);

in‐serviceinspections(mechanicalcomponentsandpiping);

erosioninspections(piping);

periodicinspections(allSSCs);

onlinemonitoring(mechanicalcomponents);

functionaltests(mechanical,electricalandI&Ccomponents);

surveillance(materialsamples);

conditionmonitoring(electricalcables);

waterchemistrymonitoring(mechanicalcomponents)


Plantprogrammesareusedboth to followupand tomaintain theoperabilityof SSCs.Ageing progresses inevitably and impairs the operability regardless of mitigatingmeasures.WhenestablishingacceptancecriteriaforprogressedageingofaSSCtheorig‐inaldesignbasisisnormallyapplied.ThisisbecausebasicallySSCsarealwaystomeetallthedesignbasisrequirementswhichhavebeensettotheminanyapplicableserviceconditions.Thenforhavingevidenceoncomplyingwiththerequirementsthefollow‐upprogrammesareaddressedandusedtoacquiredatawhichcanbecomparedtotheac‐ceptancecriteria.

Asfaruseofinternalandexternaloperatingexperienceisconcerned,TVOasamemberofWANO (World Association of Nuclear Operators) obtains information about eventsthat occur at other NPPs.WANOmembers also exchange recommendations and bestpracticesworldwidewithoneofthesubjectsbeingageingmanagementofSSCs.Similar‐ly,TVOexchangesregularlyinformationwithdedicatedSwedishworkinggroups(estab‐lishedbySwedishNPPs)forreactorpressurevesselandinternals“ReactorGroup”andturbineisland“TurbineandGeneratorGroup”.

OntheR&DsideTVOparticipatesinthenationalSAFIR(Safetyofnuclearpowerplants‐FinnishNationalResearchProgramme)thatfocusesonvariousresearchfieldspartlyre‐latedtotheageingofNPPs.TheSAFIRresearchprogrammeprovidesTVOwithaccesstovaluableresultsofseverallarge‐scaleinternationalresearchprojects.TVOmaintainsal‐socontinuouscooperationwithVTT(TechnicalResearchCentreofFinland)andFinnishtechnicaluniversities.WhennecessaryTVO'sownexpertisecanbesupplementedbyex‐pertsandresearchersfromtheseorganizations.WestinghouseSwedencoordinatestheoperationofNOG(NordiskOwnerGroup),whichoffersTVOopportunitiestojoinSwe‐dishresearchprojectsconcerningOlkiluoto1and2typeBWRs. Inaddition,TVOis in‐volved in EPROOG cooperation (EPR plants Taishan, Flamanville, Hinckley Point andOlkiluoto3).

Fortum

Ageingisassessedregularlyandreportedtotheauthorityacc.toYVLA.8requirementsineachtechnicaldiscipline;Mechanical,Electrical, I&C,CablesandCivilStructures.As‐sessmentisbasedontheresultsandfeedbackofthemainplantprogrammes:

Surveillanceandtestingprogramme;

Maintenanceprogramme;

In‐ServiceInspectionprogramme;

Chemistryprogramme.

Feedbackdata (eg. failure trends) from theplantprogrammes is analysed annually toidentifypossibleimprovementsareasoftheageingmanagementprogramme.

AconditionclassificationsystemhasbeenintroducedfortheSSCs.Itisderivedfromthefollowingcriteria

unavailability(notplanned);


averageObsoleteValueRanking;

relativechangeinfailuresandrefurbishmentsoverthepreviousthreeyears;

numberofproposedmajoroverhaulswithoutaninvestmentdecision;

numberoftemporaryrepairs;

lossofpowerproductionresultingfromaSSCfailure.

Cooperationandutilizationofoperationalexperienceinthefieldofageingmanagementtakes place through the plant programme for internal and external operating experi‐ence.

2.3.3 Monitoring,testing,samplingandinspectionactivities

TVO

TVO’s programmes for monitoring, testing, sampling and inspection activities are de‐scribedinthefollowing.

Loadmonitoring

Pressureandtemperaturetransientsoccurringintheprimaryandthesecondarycircuitaremonitoredandrecorded.Theactual transientsare thoroughlyassessedbyexpertsandthenclassifiedasdesigntransients.ThedesigntransientsaresimplifiedplanteventswhichstressSSCswiththepressure,temperatureandflowimpactsthathavebeenspeci‐fiedintheoriginalplantdesign.Loadmonitoringisimportantbecausethefatigueanal‐yses that are used to estimate the plant’s design service life have been performedthroughthedesigntransientsandexpectedamountofoccurrences.

Loadingandstresscalculations

Loadingandstresscalculationsofpipingandsupportsinsidethecontainmenthavebeenperformed and reportedwith a software PAMS. Additionally,mechanical and thermalstresses, erosion, (stress) corrosion and other relevant failure mechanisms over theplant’s service life are analysed for their effects on SSCswith PAMS. Results of PAMSanalyses are then utilized when inspection areas are selected for risk‐informed in‐service(RI‐ISI)inspectionsofpiping.

In‐serviceinspections(mechanicalcomponentsandpipingOL1,OL2,OL3)

ASMECode SectionXI is applied to theperiodic inspectionprogrammesofpipingandcomponents.Theobjectsofinspectionandthedivisionoftheobjectsintoinspectioncat‐egories are selected based on Sub‐sections IWB‐2500, IWC‐2500, R‐2500 and TablesIWB‐2500‐1, IWC‐2500‐1, R2500‐1, and the associated model drawings. The pro‐grammescovercomponentsandstructuresassignedtosafetyclasses1and2orother‐wiseassessedtobesignificanttosafety,suchaspressuretanks,pumps,valvesandtheirsupportstructures,aswellasthereactorpressurevesselinternalsandtheflywheelsoftherecirculationpumps,andpipinginspectionobjectsbasedonrisk‐informedtargeting(RI‐ISI). Changes in new revisions of ASME Code Section XI are monitored, and any

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amendmentsandchangesthatareconsiderednecessaryareaddedinthelistofinspec‐tionobjects.Theselectionprinciplesappliedtotheinspectionobjects,methodsandin‐tervalsaswellasthereportingandassessmentproceduresforinspectionresultsandde‐fectindicationsaredescribedinthesummaryprogrammeswhichprovidereferencestodetailedinspectionprogrammes.

Ageing management covers the piping included in the RI‐ISI programme. The risk‐informedinspectionmethodreferstotheutilisationofdataprovidedbyfailureanalysesandtheprobabilisticsafetyanalysis(PRA)intheselectionoftheinspectionobject.ThemethodisbasedonASMERI‐ISImethod–Boiler&PressureVesselCodeSectionXI,An‐nexR,methodB.ThemethodshallmeettherequirementslaiddownforitinGuideYVLE.5.Theriskresultingtotheplantfromapipebreakconsistsoftwofactors:thebreakprobabilityandtheadverseconsequencesofabreak.Periodicinspectionsaredesignedtominimisethisriskbyaffectingthebreakprobabilitiesofpipes.Theconsequencesofbreaks cannot be influenced by the inspection programme. The use of the inspectionprogrammesensurestheinclusionofthehighest‐riskweldsintheprogramme,aswellas the efficient use of resources. The objective of themethod is to identify the pipingsegmentsofhigh risk significance, and to select in these segments thewelds tobe in‐cludedintheinspectionprogramme.

Erosioninspections(pipingOL1,OL2,OL3)

Erosion inspectionsare targetedatareassusceptible toerosionandpitcorrosion.Theinspectionsaretargetedatpredefinedpipelines.Everyyear,anexpertgroupdrawsuptheerosioninspectionprogrammeforthefollowingyearbasedontheresultsofthepre‐viousinspections.Inaddition,pipingconnectedtopressureequipmentincludedinperi‐odicinspectionsaswellaspipelinescutoffinconnectionwithmodificationsareselectedasobjectsforvisualinspections,asfaraspossible,eveniftheyarenotpartofthemoni‐toredpipelines.Theinspectionmethodsaredeterminedbasedonthemonitoreddefecttype.Ameasurementreportbasedonthetestprocedureusedisdrawnup,andaniso‐metricdrawingwithinspectionareamarkingsisattachedtothereport.Theinspectionresultsareonanannualbasisimportedtoelectronicisometricdrawingsusedforthefol‐low‐upoferosioninspections.

Periodicinspections(allSSCs)

Some periodic inspections are regulatory requirements or they are within scheduledmaintenance programmes or particularly set to be consistwith the highmaintenancecategoryofaSSC.Theseinspectionscover

periodicinspectionsforpressureequipment; periodicinspectionsforhoistingequipment; periodicinspectionsforelectricalcomponents; periodicinspectionsforfireprotectionequipment; periodicinspectionsforcivilstructures.

Inaddition,inspectionsaremadeduringregularwalk‐downsbyoperationpersonnelac‐cording to a check list. These inspections typically include visual inspections for leak‐ages, checks for abnormal noises, lubricant level checks and vibration measurementswithaportabledevice.



Onlinemonitoring(mechanicalcomponents)

OnlinemonitoringcoversSSCsforwhichcontinuousmonitoringisconsiderednecessaryforpredictivemaintenance.AtNPPunitsOlkiluoto1and2,onlinevibrationmonitoringhasbeeninstalledtoreactorrecirculationpumpsandturbinegeneratorsets.AttheNPPunitOL3,therewillbeahigheramountofonlineconditionmonitoringsystemssuchas

loosepartsmonitoringsystemprimarycircuit; vibrationmonitoring systems for main coolant line, main coolant pump,main

feedwaterandsteamline; leakagemonitoringsystemsinsidecontainment; diagnosticofrotatingmachinery; valvemonitoringsystem; fatiguemonitoringsystem.

The continuously monitored parameters are recorded and analysed in the data pro‐cessingsystems.TheresultsareusedtoassessconditionandperformanceofaSSCcon‐cerned,andifmajorchangesareobserved,toschedulefuturemaintenanceactivities.

Functionaltests(mechanical,electricalandI&Ccomponents)

FunctionaltestsareperformedfollowingthescopesetinOperationalLimitsandCondi‐tions.Thesetestsarenormallytodemonstratetheoperabilityofstand‐bySSCsincludingemergency power supply and safety injection systems. Functional tests are also re‐quired in themaintenanceworks before putting the SSCback into service in order tohavefullconfidenceintheSSC’sfunctionalityafterrepairorpartreplacement.

Surveillance(materialspecimen)

Asirradiationembrittlementisconsideredamajordegradationmechanismofthereac‐torpressurevessel,radiatedmaterialspecimenareusedtomonitortheeffectofradia‐tion on thematerial properties of reactor pressure vessel parts in the beltline region.Thebeltlinebasematerialandweldmetalpropertiesaredeterminedthroughtheper‐formanceofmechanicaltestsforbothnon‐irradiatedandirradiatedtestspecimensforreference.Specimensarelocatedclosetothecore,areexposedtoahigherrateofradia‐tion and consequently themeasuredmaterial properties are expected to be conserva‐tive.SeealsoSection5.1.3.

Conditionmonitoring(electricalcables)

SeeSection3.1.3.

Chemistrymonitoring(mechanicalcomponents)

Chemistrymonitoringisperformedbymeansofeitherperiodicorcontinuoussamplingandanalyses.ParametersofchemistrydonotusuallyprovideimmediateinformationoftheSSCconditionsbutdeviations froma target value indicate that remedialmeasuresmaybeneededtoavoiddegradationintheSSCresultingfromthedeviation.ChemistrymonitoringcoversthefollowingSSCsandmonitoredparameters



‐ Reactorcoolantsystem

‐ NuclearIslandauxiliarysystemsincludingclosedcoolingwatersystems

‐ Water‐Steam–cycle

‐ TurbineIslandauxiliarysystemsincludingclosedcoolingwatersystems

‐ Emergencydiesels

‐ SpentFuelStoragesystems

All system’sparametersaredivided into twocategories,Controlparametersanddiag‐nosticparameters.Allparametershavetheirnormalvaluesandthreeactionlimitvalues.AlsosystemsinTurbineIslandhavesocalledtolerablevaluesasanextracategory.

Fortum

Fortum’sprogrammesformonitoring,testing,samplingandinspectionactivitiesarede‐scribedinthefollowing.

Surveillanceandtestingprogram

Periodicfunctionaltestingofthesafetysystemsisperformedacc.toTechnicalSpecifica‐tionswhich define testing of plant systems and components, operational status of theplantunit,testingfrequencyandtestprocedures.Dailysurveillanceisperformedbyop‐eratingstaffwalk‐downs.

Maintenanceprogram

Most of the condition monitoring activities are included in the maintenance pro‐grammes.

‐ on‐lineandoff‐linevibrationmonitoring

‐ motorcurrentspectramonitoring

‐ samplingandconditionmonitoringoflubricants/oils

‐ Thermographymonitoring(infrared)

Allocationofconditionmonitoringandmaintenanceactivities isbasedonmaintenancecriticalityclassification(1‐4).Conditionmonitoringisperformedonlyforcomponentsincriticalityclasses1‐3.becausethefunctionalfailureofclass1‐3componentshasanim‐pactonplantsafety.Themainpurposeoftheconditionmonitoringistodetectthedeg‐radationbeforeanyfunctionalfailureoccurs.

Chemistryprogramme

Theprogrammeofprimarychemistrycontrolisfocusedonmaintainingprimarysystemintegrityandtoreduceradiationlevelsintheprimarysystem.TheprogrammeisbasedonoriginalTechnicalSpecificationsandwaterchemistry instructionswhichreflect the



recommendationsfromfuelvendor,plantsupplierandtheexperiencesgainedatsimilarVVERunitsintheworld.

Themainpurposeofthesecondarychemistrycontrolprogrammeistoprotectthesteamgeneratorsandothercomponentsfromcorrosionrelateddamages.Itisaplantspecificprogrammesince thesecondarysideatLoviisacontainsstill somecopperpartswhichprohibittheapplicationofthehighAVT‐chemistry(projectgoing‐on,implementationaf‐ter outage2018).Otherwise the programme reflects the current understanding of theoptimizedwaterchemistryrequirementsbasedonoperatingexperienceofotherplantsandplantsuppliers.

Inspectionprogramme(ISI‐program)

Loviisa plant has ISI‐programmes for both pipelines and components. They are bothbasedonASMEXIrequirementsandarepreparedfortenyearperiods.Pipelinecondi‐tionmonitoringisapartofISI‐programmeanditismainlyfocusedonmonitoringflowacceleratedcorrosioninthesecondarycircuitoftheplant.

Otheractivitiesforconditionmonitoring

‐ Monitoringofloadsandtransientswhicharethebasisforstrengthanalysis

‐ Monitoringprogrammeforcableageing

‐ MonitoringprogrammeforRPVirradiationembrittlement

2.3.4 Preventiveandremedialactions

Thelicenseeshaveprogrammesinplacedefiningtheconditionmonitoringandmainte‐nanceofSSCsincludingschedulesfortheactionstobetaken.

Anyneedformaintenanceorrepair(SSCinserviceorstand‐by)shallbereliablydetect‐ablebymeansofconditionmonitoring(on‐linemonitoringorperiodicinspections)be‐foredegradationoftheoperabilityincursrisktonuclearsafety.Conditionmonitoringistypicallybasedonvisualinspections,non‐destructivetesting,functionaltestsandpres‐sureandleaktightnesstests.Conditionmonitoringisalsoconsideredtoincludeactionsthat provide information about parameters that are related to the SSC operability orhaveaneffectonitsuchascumulativefatigue,hydrochemistryparametersandmaterialsurveillance.WhenaSSCisrefurbishedorrepaired,thelicenseeistoinvestigatewheth‐er the samedegradation couldbe found in other similar SSCs at his plant (a commoncause failure). Moreover, the licensee is to investigate how the degradation could beavoidedinthefuturebyimprovingtheconditionmonitoringormaintenanceoftheSSC.

TheconditionmonitoringandmaintenanceprogrammesandinstructionspertainingtoaSSCarebebasedontheapplicablestandards,manufacturer’srecommendationsortheoperating experience feedback received in‐house or from other nuclear facilities. Theaimisthattheseprogrammesandinstructionsareunambiguouslyandclearlyfamiliar‐izedtotheoperationandmaintenancestaff.



2.4 ReviewandupdateoftheoverallAMP

ThelicenseesreviewandassesstheeffectivenessoftheiroverallAMPonaregularbasis.Theinputfortheassessmentistypicallygatheredfromthefollowingsources

conditionmonitoringofSSCssuchasin‐serviceinspectionsandtests;

scheduledandunscheduledmaintenanceofSSSs;

evaluationofplantspecificandothers’operatingexperiences

evaluationofageinganalysesthataretimelimited;

R&Dresultswhenapplicable;

regulator’sfeedback.

Whenevertheassessment indicatesweaknessesor impairedperformance, forexamplein termsof increasing failure frequencyofaSSC, theAMstrategy isreconsideredandmodifiedforthatparticularSSC.

TheregulatorreviewstheoverallAMPandrelatedprogrammeswhenthe licenseeap‐pliesforanoperatinglicense(neworlicenserenewal).Theassessmentofageingman‐agementissuesisalsointegratedintothereviewofPeriodicSafetyRevieweverytenthyear.

2.5 Licensee’sexperienceofapplicationoftheoverallAMP

TVO

Normalway tomanage any technical problem is to consider some corrective actions.IdeaofallTVO’sAMPsisbytestingormonitoringSCCstofindoutdeficiencies.Correc‐tiveactionshavetobestartedifadeficiencyisindicated.Thisisanormalactiontomodi‐fyperiodicorpreventivemaintenanceorchangetheactionitself.Thepipelinemanage‐mentprogrammeRI‐ISI,orErosion‐programmeetc.isupdatedbasedonthefindings.

Sometimes amore detailed AMP is seen necessary. TVOwill produce somemore de‐tailedAMPsin2018,partlyduetostartupofOlkiluoto3.

Fortum

Continuous improvementhasbeendoneover theyears in the fieldof ageingmanage‐mentinLoviisaplant.

In2002majorre‐organizingwasmadeandageingmanagementrolesandresponsibili‐tieswereclarifiedatthattime.ThescopeofAMPwasalsoclarified.Mainreasonforthatwastheupcominglicenserenewaloftheplant.Thenewlicensewasgrantedin2007andsince then the ageingmanagementorganization, processes and scopehavebeenquitestableuntil2015.



In2015GuideYVLA.8waspublishedand thescopeof theAMPhad tobeadjusted tomeetthenewrequirements.Somenewplant levelguidelinesandprocedureswerede‐velopedtosupporttheeffectiveageingmanagement.Thisworkisundercontinuousde‐velopment.

Duringthe40yearsofplantoperationtherehasbeensomeunexpectedageingrelatedincidents,e.g.themostsevereweretwofeedwaterpiperupturesin1990's.Wheneverthiskindofincidentshaveoccurredthecorrectiveactionshavebeentakenbasedonthecase‐specific lessons learned. In practice the scope of AMP or AM‐related plant pro‐grammemodificationhavebeenatypicalcorrectiveaction.

AtthemomentthescopeofAMPandplantlevelAM‐relatedprogrammesareadequatetoensurethesafeandreliableoperationofthepowerplant.

2.6 Regulatoryoversightprocess

The regulatory oversight of ageing in FinnishNPPs focuses on review of ageingman‐agementprogrammesalongwithoperating licenseapplicationsandperiodicsafetyre‐views(PSRs)wheretheconformancetotherelevantSTUKRegulationsandYVLGuides,including experiences in licensee’s recent ageingmanagement, is investigated. STUK’sfindingsfromotherregulatorycontrolpracticesareusedasverification.

Theperiodicinspectionsareperformedonplantsiteaccordingtoannualplanningandtheytackleboththeorganizationalissuesandtechnicalaspectsofeachdiscipline.With‐in the periodic inspection programme there is one dedicated inspection, called PlantMaintenance,whichexclusively concentrateson the conditionmonitoringandmainte‐nanceactivitiesandageingmanagement.Theaimof this inspection is to evaluate andverifytheproceduresthelicenseehasforensuringreliableintegrityandperformanceofSSCs.STUKwillalsoassesstheimplementationoftheageingmanagementprogrammesbasedonthefollow‐upreportspreparedannuallybythelicensees.

AnexpertgroupdedicatedtoageingmanagementhasbeenestablishedwithinSTUKtooverseehowthelicenseesperformtheirdutiesintheageingmanagementofSSCs.Thegroup,which consists ofmechanical, electrical, I&C, civil and human resource expertsand resident inspectors, plans and coordinates STUK’s regulatory duties pertaining totheageingissuesofFinnishNPPs.Oneofthemajortasksistoevaluateageingmanage‐mentprogrammesandannualfollow‐upreports.Ifshortcomingsarefound,forexampleinattendingtothemaintenanceofaSSCinthelongterm,thegroupcallsthelicenseeforfurtherclarificationsorpossiblecorrectiveactions.Thegroupalso followsup findingsfrom other countries and evaluates their possible applicability to the ageingmanage‐mentoftheFinnishNPPs.

2.7 Regulator’sassessmentoftheoverallageingmanagementprogrammeandconclusions

Aregulatoryguideforageingmanagement,GuideYVLA.8hasbeenrecentlyissuedandenforced. In the Guide, there is a requirement for the licensees to draw up an ageingmanagementprogramme for regulator’s approval.Themain topicsof the ageingman‐agementprogrammearesummarizedbelow



Coordination, responsibilities and duties (organizational issues of ageingman‐agement);

Measurement of effectiveness (evaluation of the performance of ageing man‐agement);

Utilizationofoperationalexperienceandresearchdata;

Gradedapproach(whenappliedtoageingmanagement);

DataofeachSSCorcommoditygroupwithinlicensee’sageingmanagement(de‐sign basis, ageing mechanisms, condition monitoring and maintenance pro‐grammes,specificAMPs,TLAAs.);

Provisionsformanagementofobsolescence.

Inthebeginningof2017TVOandFortumissuedtheirupdatedageingmanagementpro‐grammesaccordingtoGuideYVLA.8.Basedonthereviewof theseprogrammesSTUKconcludedthattheybothstillneedfurtherdevelopmenttosomeextent.DeadlinefortherevisedageingmanagementprogrammeswassettotheendofApril2018.Regardlessofdeviations from thenewregulatory requirementsboth licenseeshavehadsatisfactoryageingmanagement approaches since the commissioning of the plant units. However,theroleofcomprehensiveageingmanagementprogrammeswillbemoreemphasizedastheoperationoftheNPPunits isextendedbeyondtheoriginaldesignlifetime.Thede‐signbasisoperabilityofSSCshas tobemaintainedeventhoughtheirdegradationratemaybehardtoanticipate.

AgenericlessonlearnedinFinlandisthattheclosernuclearpowerplantsgettotheendoftheirdesignlifetime,themorechallengingitisforthelicenseestostartlargeandex‐pensiveinvestmentstomoderniseormodifytheNPPs.

Insteadofrenewingasystemoracomponent,modernisationmaybepostponedorreal‐izedonlypartially.ApostponeddecisiontorenewforinstanceanI&Csystemoranelec‐tricalsystemmayresultinanobsolescenceofsystems,i.e.,sparepartsortechnicalsup‐portarenolongeravailable.Thismayleadtosituationswherethelicenseemaynotbeable todemonstrate thesafetyofoperations to theregulator,oras faras thescopeoradequacyofdemonstrationisconcerned,opinionsmaydifferbetweenthelicenseeandtheregulator.Finlandhassuccessfullyappliedperiodicsafetyreviews(PSR)fortheop‐eratingNPPs.ThelicenseesareobligedtodemonstratethatthesafetyoftheoperationscanbeensuredandimprovedalsoduringthetimebeforethenextPSR.Inasimilarway,theyhavetocommittocontinuoussafetyimprovementsintermsofmodernizationpro‐jectsinordertomanagebothphysicalandtechnologicalageinginthelongterm.



3 Electricalcables

3.1 Descriptionofageingmanagementprogrammesforelectricalcables

NPPunitsLoviisa1and2:

Theageingmanagementprogramme(AMP)forelectricalcablesinLoviisanuclearpow‐erplantinvolvesallthecablesandcabletypesusedinLoviisaNPP.TheAMPmonitorsthecablesconditionsindifferentbuildingsandatdifferentenvironments.FortheAMPofcablesnotonlythecablesareincluded.Alsothecableways(e.g.trays),connections,joints,terminalsareincludedaswellasthestructureswherethecablesortheirwaysareincontact(e.g.concretestructures).TheAMisongoingprogrammeandthereportingtoFinnishRadiationandNuclearSafetyAuthorityisdoneeveryfouryears.

NPPunitsOlkiluoto1,2and3:

TheageingmanagementofelectricalcablesofOL1andOL2isdefinedinthegeneralage‐ingmanagementprogrammeofTVO’snuclear facilitiesOL1,OL2,OL3andKPA:Docu‐ment117279.TheprogrammedescribestheageingmanagementprocessfortheSCCsofNPPunitsOlkiluoto1,2and3andthespentfuelstorage.Howeverthemoredetailedfa‐cilityspecificprogrammesarefocusedonSSCsimportanttosafety.Thefocusingisbasedonthetechnicalspecifications,safetyclassificationandPRA.

Related to the general objectives of the ageing management (such as organization,choosingbases,analyses,qualificationmanagementandhandlingoftheresults)theage‐ing management programme of Olkiluoto do not handle cables separately. The cablesamplingprogrammeofOlkiluoto1and2containmentcablesandplans forOL3cablesamplingarereferredinthegeneralprogramme.

3.1.1 Scopeofageingmanagementforelectricalcables


InLoviisaNPPthescopeofAMPforcablesareallMV(20kVand6kV),LV(<400V)andI&C(24VDC,48VDC,220VDC)cables.AlsospecialcableslikeLOCA,mineral,thermal,coaxialandfibreopticcablesaswellascablejointsandterminalsaretakenintoaccountofAMPinLoviisaNPP.PriorityofAMPforcablesliesinthesafetyrelatedcablesthatareinthemostharshenvironmentandunreachableduringtheoperation,i.e.onlyreachableduringoutages,mainlythesteamgeneratorroominthecontainment.

CurrentlytheusedLOCAcablesinthesteamgeneratorroomarethepowerandI&Cca‐bles fromSiemens (Sienopyr) andHabia. The other safety related cables in the steamgeneratorroomarefromNokia,AcomeandThermocoax.Inthesteamgeneratorroomliesalsocabletypes(MHMS‐Si,MonetteandSSJS)thatdon'thavesafetyfunctionbutareneededforoperation.Theirconditionisalsofollowedwithsamples.

Other cable typesused inLoviisaNPPare frommanufactures such asReka, PrysmianandHelkama. It isrecommendedtouse theHalogen‐Freecable types fornewinstalla‐tionsandrepairs.



Thesafetyclassifiedcable’suitabilityshownbyanalysisandtests.Ifpossible,thecablechosenforqualification,itisreasonabletoqualifythecabletypeforallthesafetyclassesandavastrangeofenvironmentalconditions.Alsoifthewholecabletype(meaningallitscablesizes)isqualifiedtherightcablesizeshouldbeavailableandthereforetheelec‐tricalsuitabilityismucheasiertoprovefortheinstallations.Insomeplacesitispossibletoonlyqualifyaspecialcabletype,likeLOCA‐cable,sothevarietycan'tbethatlarge.

The recognized ageing mechanisms for cables are: high temperature, short circuit orovercurrent,voltageload,ionizationandUVradiation,lackofcoolingand/orventilation,mechanicaldamagesandinappropriatesubstancesgettinginvolvedwithcable.

Theabovementionedcanbediscoveredwithvisualinspectinglikechangeinthecolorofthecable,crackinthecables'jacketetc.Alsoinspectionslikethermalmeasuringareim‐portant. Sampling is also done for the cables in the steam generator room. Arrheniusmethodisusedwhentheestimatedlifetimeofthecableisneeded.

NPPunitsOlkiluoto1and2:

TheageingmanagementofelectricalcablesinNPPunitsOlkiluoto1and2concentratesonthenuclearsafetyclassified(SC2andSC3)cablesinmostdemandingnormaloperat‐ingserviceenvironmentsandtheageingmanagementproceduresaremorerigorousforthecableswhichinadditionarerequiredtooperateunderharshenvironmentsindesignbasisaccident(DBA)conditions(i.e.lossofcoolantaccident,LOCA).

ThescopeforconditionmonitoringofelectricalcablesofNPPunitsOlkiluoto1and2isdefinedintheinstruction108654.

As for the in‐containment cables, the scope of conditionmonitoring covers the safetyclassified(SC2andSC3)cabletypeswhichareinuseinallareasofthereactorcontain‐mentbuildings(lowerdrywell(LDW),wetwell(WW)andupperdrywell(UDW))ofOL1andOL2.Thecablesamplesforconditionmonitoringtestsarelocatedinmostdemand‐ing normal operating service environments, primarily inUDW, in locationswhere thetemperature and radiation levels are the highest. Cable loops for resistancemeasure‐mentsareinstalledalsoinWW.Atpresentthesamplecabletestsoftheconditionmoni‐toringprogrammecoverthefollowing“in‐containmentgrade”cable families,whichin‐cludelowvoltagepower(<1kV),controlandinstrumentationcabletypes:

Lipaloncables(LiljeholmensKabelfabrikAB/AseaKabelAB)

FirewallIIIcables(TheRockbestosCompany)

“OL1/OL2LOCA”cables(HabiaCableAB).

Similartestsandinspectionsareperformedalsoonselectedoutside‐containmentcabletypes,whichingeneralareFinnishnational,PVC‐insulatedand‐sheathed,standardca‐bletypesforlowvoltagepower(<1kV),controlandinstrumentationinstallations(e.g.MMJ,MMO,MCMK,MMAO).

In addition, e.g. partial discharge (PD) measurements of non‐nuclear safety classifiedmedium voltage cables (6,6 kV system) important for plant operation have been per‐



formedaspartofprojectsandasseparatecableconditionmonitoringcampaignstopre‐ventserviceinterruptionsandensureplantavailability.

The ageing stressors and degradation mechanisms applicable for electrical and I&Cequipmentare identified inTVO’sTechnicalRequirementsand Inspection Instructionsforenvironmentalqualification.Thesignificanceoftheageingstressorsanddegradationmechanisms for cables is evaluated in suitability analysesof the cables. The conditionmonitoringactivitiesdescribedaboveconcentrateonthecables’insulationandsheath‐ingstructuresmanufacturedoforganicpolymermaterials,whicharemostsensitivetothermal and radiation induced ageingdegradation.The loop resistancemeasurementscoveralsotheconductormaterial(copper).

Test results and experiences from inspections and sample cable tests are utilized inevaluatingtheconditionofthecablingintheplantsgenerally.

The ageing management activities have been started in 1982. Processes, procedures,methodsandcriteriaetc.havenotbeenverypreciselydocumented,butapparentlyallsignificantaspectshavebeenconsidered.

NPPunitOlkiluoto3:

The conditions affecting the ageing mechanism of electrical cables are identified andevaluatedintechnicaldocuments,suchasProjectSpecifications,QualificationSpecifica‐tionsandSuitabilityAnalyses.Theconditionsinclude:

ambienttemperature

temperatureriseduetoconductorloading

voltagestressininsulation

humidity

mechanicalstressandvibration

UVradiation

radioactiveradiation.

Specialattentioninthiscontextisgiventoeffectsofradioactiveradiationtogetherwithother factors, assuming that engineers and cable manufacturers have enormousknowledge and long‐time experience of ageing of different cable types from otherbranchesofindustryandutilities.

InOL3,thepreliminaryapproachtodefinethescopeofageingmanagementistoincludeall cables feeding or controlling equipmentwith safety functions and thus assigned tosafetyclassSC2orSC3(quantityofcablesapprox.16000ofthetotalofapprox.47000cables inNI). In addition the scope of ageingmanagement shall include cables havingimportantroleinnuclearsafetyfunctionsandcablesinstalledinexceptionallyharshen‐vironment,aswellascablesconsideredsignificantforplantavailability.



Cableswithinthescopeofageingmanagementaretreatedindifferentcategoriesasfol‐lows:

10kV(inIECterminologyMediumVoltage)cablesforRCPmotorfeeders

<1kV(inIECterminologyLowVoltage)powerandcontrolcables

I&Ccables

Specialcablese.g.forECI,RPVLandN16instrumentationcables

Opticalcables

Coaxialcables.

For ageingmanagement purposes the cables will be treated in groups, whichwill beformedaccordingtocategory,ambientconditions,cabletype,materialetc.Groupinghasnotbeendefinedyet,butitisexpectedthatapproximately50groupswillbeformed.

Todefinewhichcabletypesshallbeinstalledincabledeposits,thefollowingprocedurehasbeenused:

extract data of cables and cable routes from theKADISdatabase (which is theprojecttoolformanagingcabledataandrouting)

identifyfromtheRoomConditionsReportallrooms(insidethecontainmentandout‐side)whereRoomRadioactiveDoserateis2mSv/horhigher(socalledredrooms)

recognize fromthetotalcable listofNIallcableswhichhaveoriginordestina‐tioninsuch ‘red’roomsorareroutedthroughsuchrooms(totalapprox.3600cables)

frompreviousselection identifycableswhichbelongtosafetyclassSC2orSC3(totalapprox.1000cables)

consolidatelistcountingquantitiesofdistinctcabletypes(61types).

3.1.2 Ageingassessmentofelectricalcables


AtLoviisaNPPtheageingassessmentofelectricalcablesisbasedonIAEATecDoc1188anddocumentNP‐T‐3.6.InqualificationthestandardsusedareIEEE308,IEE383,IEEE323andIEC60780.Withthehelpofthesestandardsthequalificationisdonebyemulat‐ingtheartificialageingofthecable.Alsotheregulatoryguides[YVLA.8]and[YVLE.7]togetherwithmaintenanceinstructionsarefollowed.



Fortumhasbeenapartand/orstudiedofmultipleprogrammeswithothercompanieslikeVTT(TechnicalResearchCentreofFinlandLtd),AaltoUniversityandElforsk(Swe‐dish Energy Research Centre)with programmes like KaLiFi, POSSE and SAFIR. KaLiFiprogrammewasdonewithAaltoUniversityandithandlestheageingoftheMVcables,POSSEwasdonebyVTTandithandlesthePolymers.SAFIRisalargeprogrammethathasbeendoneincollaborationbySTUK,Fortum,TVO,VTT,KTM(MinistryoftradeandIndustry),Tekes(FinnishFundingAgencyforTechnologyandInnovation),TKK(Helsin‐ki University of Technology) and LUT (Lappeenranta University of Technology). TheSAFIRprogrammehandlestheNuclearPowerPlantsafety.

Fortumisalsopartof thecollaborationofothernuclearplants inScandinavia.Fortumhasbeenapartofseminarsasanorganizerandasaparticipant.AlsotheotherNuclearOperatorsandothereventsintheworldarefollowed.Iftheeventscanbesomehowre‐lated to LoviisaNP the eventwill be studied and itwill be estimated if the event canhappeninLoviisa.Ifitcansomekindofactionswillfollow.

Someof the results from theseprogrammesand collaborationshasbeen taken inuse.ForexamplewithcollaborationwiththeSwedishwasfoundthatAcome1Ecabledoesn'tfunctioninLOCAasitwassupposedtoandalsotheHabiacabledoesn'tfunctionwhensunkunderwater.


Currently there is no AgeingManagement Programme or documented assessment onageingofelectricalcables.PreparationofAMPisunderway.Genericinternalinstructionsofageingmanagementapply.

KeytechnicaldocumentationofNPPunitsOlkiluoto1and2cablingisshowninAnnex2.

TVOhasparticipatedseveralnationalR&Dprogrammesregardingcables.Themostre‐cent re‐searchprojects related to ageing and conditionmonitoringof cables are listedbelow:

COMRADE (ConditionMonitoring, Thermal andRadiationDegradation of Poly‐mersInsideNPPContainments,2016‐2017)

FIRED(Ageingofflameretardantcables,2016‐2017)

POSSE(Studyonevaluationofradiationresistanceandbasesofinspectabilityofpolymers,2016)

KaLiFi(Developingconditionevaluationanddetermining lifetimeof intermedi‐atevoltagecablesbyelectricalandchemicalmethods,2005‐2007).

TVOhasalsobeenactive intheSwedishNuclearPowerPlantcompanies’co‐operationgroup “EKGKabelgruppen”. The co‐operation gives valuable experiences and researchdatafromotherplantsusingsametypeoftechnology.



NPPunitOlkiluoto3:

PreparationofAgeingManagementProgrammeforelectricalcablesiscurrentlyunder‐way.

Ingeneral,theprinciplesandrecommendationsofIAEAdocumentNPT3.6“AssessingandManagingCableAgeinginNuclearPowerPlants”isfollowedwhenapplicable.

Standardsrelevanttocableageing:

IEC60544 Determinationoftheeffectsofionizingradiation–Proceduresforassessingageinginservice

IEC62465 Managementofageingofelectricalcablingsystems

IEC62582‐3 Electricalequipmentconditionmonitoringmethods–Elongationatbreak

KeytechnicalOL3projectdocumentationrelevanttocablingareshowninAnnex3.

3.1.3 Monitoring,testing,samplingandinspectionactivitiesforelectricalcables


There are instructions inwhich the Loviisa NPP buildings are categorized so that foreverybuildings'electricalinstallation(incl.cables)themonitoring,testing,samplingandinspections isdoneperiodically.Mostly themonitoringand inspectionsarevisual, butsomeactionse.g.forcingaloadtoacabletraysmightbedone.

The visual monitoring and inspection cover the condition of cables and trays, joints,connections, groundings, fireprotection, possiblemechanical damages, inspection thatthecablesareontherighttrayetc.Specialattentionmustbetakenforinstallationsthatare in a humid environment. Every year is done the visual inspection for the thermalprotectionofthecablesinthesteamgeneratorroom.Radiationlevelofthesteamgener‐atorroomisfollowedaswell.

For theMV cables someextraperiodical testing is alsodonee.g. insulation resistance,tan‐deltaandoff‐linePD‐tests.

Inthesteamgeneratorroomthereareathreecabledepositswheresamplesaretaken.Inthesedepositsaresituated0,5msamplesofthecabletypesusedinthesteamgenera‐tor room. Two of these deposits are in normal operation environment and one is inharshconditions.Samplesaretakenperiodically(every4years).Thesamplesaresenttoanoutside company.Thecompanyperforms tests toeach sample, elongationatbreakandtensilestrengthatbreak.



Thesameyearwhenthesamplesaretakenfromthesteamgeneratorroomisalsodonethermalmeasurementforcabletraysandsinglecables.Thethermalmeasurementfromconcretestructuresinthesteamgeneratorroomisdoneeveryyear.Theresultsaredoc‐umentedandalsothesourceoftheheatshouldbedocumented.

Determining if thecable is longervalid for itsuse iscase‐specific.Forobviousreasonsthemoreimportanttosafetyorproductionthecableisthemoreimportantitistokeepit functional.Thecablesamplesthatgothroughtestingwherethetensilestrengthandthe elongation at break aremeasured. If the elongation at break value is higher than50%thecableisconsideredtobesuitableforuse.Iftheelongationatbreakvaluebeginstoreachthe50%limitactionsmustbeconsidered/taken.

Thedata isgatheredfromthemonitoring, testing,samplingandinspectionsandsavedforalateruse.Afterwardsthedatacanbeusedwhene.g.informationaboutthecableshistoryisinspected.Somedatawillbeneverusedbutsomeisvaluableallthetime,eg.theradiationlevelandthethermalmeasuringareveryimportantinformationwhenfol‐lowingthetrendsofsteamgeneratorroom.Thatkindofinformationcanbeusedwhennewcablesarequalified.


Visual and tactile inspections are performed for cables, conductors, connections andterminalboxesinsidethecontainment.Theinspectionswillbedoneduringmaintenanceoutageseverysecondyearsothatapproximatelyonethirdoftheobjectsinthescopeofinspectionsareinspectedduringoneoutage.Inspectionintervalforeachobjectwillthusbeapproximatelysixyears.Theinspectionsconcentrateoneffectsofharshenvironmentand vicinity of hot objects and signs suggesting ageing phenomenon. The results aredocumentedandanalysedbyexpertsfollowingdecisionsonfurtheraction.

Therearealsospecifictestcables(typeHHSO3x1)inthecontainmentofeachplantunit,installed for thepurposeofmonitoring theperformanceof cables. Conductor loop re‐sistanceandinsulationresistanceismeasuredandrecordedonceayear.

Lipalon cable sample deposits are installed in six locations of each plant unit. The Li‐palon cable samples have been in the containments since the commissioning of theplants. Installed cable types are HHO 3x1, HHO 4x1, HHSO 3x1 and HHSO 4x1. Cablesamples(lengthapprox.1m)willbetakenatfiveyear’sintervals.Inaddition,thereareninesamplesoneachplantunitofHabiacabletypeOL1/OL2LOCA1x(3x1)/screen,in‐stalled year 2011 and one sample of Rockbestos Firewall III cable type RXSR‐G3×AWG16inOL1installedin1992.

Sampleshavealsobeentakenfromcableswhichhavebeeninuseoutsidecontainmentanddismantledforthereasonofsomechangework.

The samples are tested in an appropriate laboratory. Tests include tensile strength,elongationatbreak, voltage test and insulation resistancemeasurements.The instruc‐tionreferstostandardsIEC811andSFS‐IEC855.Thegeneralacceptancecriteriaforthesamplecabletestsaresetasfollows:



Elongationatbreak:>50%absolute(generalacceptancecriterion).Inaddition,forLOCA‐cablesthetensiletestresultsof theLOCA‐testsprogrammeare takenintoconsideration.

Voltagetests:Nodielectricbreakdownallowed.

Insulationresistance:Referencevaluesaccordingtocabletypespecificstandardsand/orTVO’sinternalinstructionsandelectricalsafetyregulations.

Inaddition,thetestresultsarecomparedwiththeresultsoftheprevioustestsfortrend‐ingandactionswillbetakenbasedonanalysisbyexperts.

NPPunitOlkiluoto3:

Procedures forregular inspectionsforcables inusehavenotyetbeendefined. Inspec‐tionswillincludevisualandprobinginspections,measurementsandPDtestsfor10kVcables.

Cablesampledepositswillbeinstalledinvariouslocationsoftheplant.Exactlocationshavenotyetbeendefined,but the target is to findconditionsrespectingtheactual in‐stallationconditionsof therespectivecable type.Thecablesamplesmaybeeitherap‐prox.35msinglelengthsor10x3,5mreadycutlengths.Cableendsshallbesealedinei‐thercasetopreventpotentiallocaldeteriorationoftheinsulationmaterial.

Preliminarylistofcabletypestobeinstalledindeposits:

N2XSH10kV1x240RM/35

N2XH‐J1kV3x2,5RE

HXELCHXOE1kV3G2,5

JE‐HCH4x2x0,5BDSI

JE‐LiHCH2x2x0,5BDSI

NHXCHX1kV3x6RE/6FRNC‐BX

(N)HXSCHXOE‐J1kV4x2,5FRNC‐BX

SISIF500V1x16FRN*S

SIHGLCSI500V4x2x1,5FRNC*S

SuperscreenedCoax21AWGHFI150–50Ohm

MIThermocableTypeK

LNCoax24AWG50Ohm

JE‐LiHXCHX2x2x0,5BDSIFRNC‐BX



JE‐LiHXCHX1x4x0,5BDSIFRNC‐BXEUPEN

JE‐HXCHX20x2x0,8BDFRNC‐BX

GOREGSC‐02‐25198‐00.

Cablesamplesshallbeextractedfromthedepositsatapprox.fiveyearsintervalsafteraninitialperiodoftenyears.

Teststobeperformedforeachsample:

VoltagetestforPowerCablesaccordingtorelevantproductstandards

ConductorloopresistancetestforI&Ccables

Insulationresistanceadjustedtosamplelength(volumeresistivity),testvoltageaccordingtorelevantproductstandards

Tensile test with elongation at break recording according to IEC 60811 orIEC/IEEE62582‐3.

Sametestsshallbeperformedtoun‐agedcablesamplesatthebeginningofdepositperi‐od,tocollectreferencevaluesoftests.

Acceptancecriteriafortestresults:

PDmeasurement:NodetectablePDat2xU0

Voltagetest:Passed

Conductorloopresistance:Accordingtorelevantdatasheetorproductstandard

Volumeresistivity:According to relevantproduct standards, if available, orac‐cording to IEC 60885‐1 (withdrawn standard), also generic value of ≥108Ωmmaybeusedasreference

Elongationatbreak:50%absoluteorrelativeindicatestheendoflife;otherval‐uesmaybejustifiedforspecificcompoundmaterials.

Multiplesamplesmaybeappliedifthereisanindicationofdispersionintestresults.Al‐somoreconservativealertcriteriawillbedefinedtoinitiatepreventiveactionintime.

The elongation at breakduring a tensile test is thebenchmark property bywhich thestructuralintegrityofthecableisassessed.Thepurposeoftheelectricaltestsdescribedaboveistoverifythatthesampleisnotdamagedandisinoperationcondition.



3.1.4 Preventiveandremedialactionsforelectricalcables


Shouldthemonitoring, testing,samplingandinspectionsshowresultsthatthecable isormaybeinbadconditionpreventive/remedialactionsmustbeconsidered.Theactionsmightbeasfollows:morefrequentmonitoringofthecablescondition,renewingtheca‐bleorassemblingacablejointiftherenewalisveryhardtomake.Inthesteamgenera‐torroomthecablejointsareavoided.

If there isaminordamage in thecable (e.g.minorviolationof jacket) then temporaryrepair should be taken, e.g. securing the crackwith electrical tape. Or if a cable traysconnection is loose, it should be tightened. After the temporary or a little repair thedamageand/oractionshouldbereportedandthenmaketheconsiderationsifafutureactionsisnecessaryorifnot.

Ifthesamplestested(3.1.3)showthatthecabletypemightbevalidonlyforashortpe‐riodoftime,thentheactionofanewcableinstallationshouldbeconsidered.Especiallyifthecableissafetyrelatedandthecableisnolongeravailable(inthestorageorbythemanufacturer), then must be done a qualification for some other type of cable. Alsorequalificationofcabletypemightbepossible.

TherequalificationwasdoneforSiemensSienopyrpowerandI&Ccables.Thequalifica‐tionofthesecableswascomingtoanend(30afromthefirstinstallations).Thecablescouldhavebeenreplacedtosomeothertypebutitwouldhaveneededalotofresourcesand time because the new cable type must have been qualified for LOCA conditionswhich involvesarisk that the firstcable typemightnotbevalidso thatasecondtypemustbequalified.Thereplacementwouldhavealsorisentheradiationlevelsoftheem‐ployersbecauseworkwouldhavetakentensofhoursinthesteamgeneratorroom.Andofcoursethereplacingallthecableswouldhavebeenveryexpensive.TheSienopyrca‐blesseemedtobeingoodconditionsotherequalificationcameasubject.TheageingandLOCAtestsweredonetothecablesandtheresultsshowedthatboththepowerandtheI&Ccablecanbeusedforover50years.

As an example of more bigger action done is the steam generator room cable routechangewhichwasdonebecauseof theresults from inspections. Itwasnoticed that insomeplacesthecableswereageingfasterthanexpected.Forpreventiveactionventila‐tionandcoolingwereimprovedandthecableroutesweremovedinaplacewherethetemperature and/or radiation is smaller. Also some obstacleswere installed betweenthecableroutesandtheheatsources.

Otherexampleofpreventive/remedialactions is theNokiaLJNSMcablerenewing.TheNokiaLJNSMcablesoutersheathwasfoundtobeinbadcondition.BecauseoftheAMPforcables theconditionof thecableswasdiscoveredearlyenough toperformactions.Thediscoverywasdonebyvisualinspectionsandmechanicaltests.Thecolouroftheca‐blesoutersheathhadchangedfromgreytodarkbrownandthejacketwasverystiffandfragile.NoticethattheLJNSMs'insulationwasingoodcondition.Becausethetempera‐tureinLO2islowerthaninLO1,theLJNSMcableswereinworseconditioninLO1thaninLO2.ForthatreasonalltheLJNSMcableswerechangedinLO1afterthediscoveryofthebadconditionoftheoutersheath.




Theinspectionandtestresultsareanalysedbyateamofelectricalexpertscasebycase.In case of tests results indicating significant degradation, sample collection intervalcould be shortened and/or additional test methods, including repeating some of theoriginalqualificationtests,couldbetakenintouse.Untilnowageing‐relatedfactorshavenotinitiatedmajorchangesinthecabling.Allcorrectivecablereplacementactionswillbeexecutedfollowinggenericinstructionsgoverningplantchangeengineeringandim‐plementation.

NPPunitOlkiluoto3:

If the test results indicate significant degradation, possible actions are shortening thesamplecollectionintervalortoperformfullLOCAtestsequencetothesampletohaveamoreaccuratepredictionofremaininglifetime.

Temperatureandradiationwillbemonitoredandrecordedintheplantgenerallyandatthe locations of deposits. History conditions will be verified referring to conditionsspecifiedtoqualificationofcables.

3.2 Licensee’sexperienceoftheapplicationofAMPsforelectricalcables


Theageingmechanismsforelectriccableshavegonemostlythewayitwaspredictedoranalysed.Somecablesseemtobeevenmorehealthythanitwasfirstanalysed(requali‐ficationofSiemensSienopyrCables,Section3.1.4).Somecabletypesseemtobeageinginawaythattheyprobablywon't lastthewholeoperationlifeofLoviisa.Thesecableshave functionedmore than Loviisa NPP originally planned lifetime (30 a, now 50 a).Thesecablesaresituatedinsteamgeneratorroomandarenotsafetyclassified,exceptNokiaLJNSM(SC3).Themainreason forageingseemtobe the temperature.Thehightemperaturesweakensthecablesoutersheathandmakesitfragile.Theconductorsandtheinsulationofthecableshavemostlybeeningoodcondition.

TheAMPofcablesinthesteamgeneratorroomwasoriginallybasedmainlyonmainte‐nanceworkdoneduringtheoutage.Cablesamplesweretakenandanalysesweremadeeveryfiveyears.Andeverysixthyearamoreprofoundvisualinspectionwasdone.De‐spitethecontinuousmonitoringitwasdiscoveredinmiddle90'sthatanumberofcableswerenotingoodcondition.Mainlybecausetheroomtemperaturewashigherthanspec‐ifiedandtherewerehotspotsinsomepartofthecableroutes.

ForthosereasonsthefollowinghasbeendonetotheAMPofcables:

Moreefforthasbeenputontheenvironmentalconditionmonitoring

Moretemperaturesensorshavebeeninstalledforcontinuousmonitoring

Profoundmanual temperaturemonitoringonoutagewith infraredpointmeas‐urementsandthermographiccameraaretakenevery4years.



Theradiationlevelismeasured

Thedepositswherethecablesamplesaretakenhavebeeninstalled

Visualinspectionsaredonenowevery4yearsinsteadof6

Knowledge has risen where the visual inspections should be done more pro‐foundly.

Nota single failure causedby the cabledegradation in the steamgenerator roomwasrecognized.Usuallycablefailuresarecausedbysomeotherworkwherethecablegetsmechanicaldamage.TheexamplesinSection3.1.4showthattheAMPforcablesiswork‐ingproperly inLoviisaNPP.Thediscoveriesof cables inbad conditionhasbeendoneearlyenoughandtheactionshavebeentakenfastenoughsothattheallthecableshavebeenfunctionalbeforetheywerechanged.


A considerable amount of data has accumulated from the tests and inspections per‐formedduringthepastyears.Thismaterialisusedasreferencewhenevaluatingfutureactions.

TVO'sexperienceson the in‐containmentcable typesofOL1andOL2hasmainlybeengoodandmajorpartofthesecablesareknowntobeingoodcondition.Mostofthede‐fectsrelatedtothein‐containmentcableshavebeenmechanicaldamagecausede.g.byplantmodificationworknearthecables.Thedamageshavenormallybeenlimitedtotheouterjacket/screenlayersofthecables.Electricalfailuresofin‐containmentcableshavebeen very rare. However, as part of component renewals of ELMA project (LifetimeManagementofElectricalandI&CComponentsInsidetheContainment)alsorenewalsofin‐containment cables with long‐term LOCA‐requirements have been ongoing since2011andaccordingtothepresentschedule,willbecompletedin2019.

AlsotheoutsidecontainmentcableshaveproventobeveryreliableduringtheoperatinghistoryofOL1andOL2.Electricalfailuresofcableshavebeenrare.Mostofthecablede‐fectsandfailureshavebeenmechanicaldamagecausedbyplantmodificationworknearthecables(e.g.cablingworkonexistingcableroutes).CablepenetrationsoftypeMCTorconsistingoffirestopmortararetypicallocationsforthiskindofdefects.AgeingrelatedcabledefectsinOL1andOL2havebeenlimitedtocasesinwhichthedesigntemperatureofacablehasbeenexceedede.g.incaseofconnectiontoawarmobject(i.e.designer‐rors).

In addition,most of themediumvoltage cables (6,6 kV system) ofOL1 andOL2 havebeenreplacedbetween2005and2013aspartofswitchgearrenewalprojectsandlaterbecauseofinstallationdefectsfoundintheterminationswithPDtests.Alsothe110kVoil‐filledcableshavebeenreplacedwithXLPE‐insulatedcablesin2008‐2010,becauseofobsoletetechnologyandavailabilityissuesconcerningrepairpartsandlackofinstal‐lationexpertise.



NPPunitOlkiluoto3:

PreparationofAgeingManagementProgrammeforelectricalcablesiscurrentlyunder‐way.NoexperiencegatheredsofaroftheapplicationofAMP.

3.3 Regulator’sassessmentandconclusionsonageingmanagementofelectricalcables

STUK’sregulatoryguideYVLE.7forelectricalandI&Cequipmentrequiresthatalicen‐seehasaplanforthemonitoringoftheageingofthecablesinsidethecontainment.Age‐ingofother cables important to safety ismanagedby thegeneral ageingmanagementprogrammesaccordingtotheSTUKguideYVLA.8forageingmanagement.

STUKhasinspectedtheageingmanagementprogrammesincludingthecontainmentca‐ble ageing monitoring plans. Implementation of the cable ageing management pro‐grammeshasbeenverifiedduringtheSTUKinspections.

The cable ageing management processes described in the ageing management pro‐grammeshavebeenfoundsatisfactorybuttheageingmanagementprogrammeofcablesonNPPunitOlkiluoto3(underconstruction)isstillunderpreparation.

Theageingmanagementprocess includes that the licensee regularly evaluate theade‐quacyofcontainmentcableageingmonitoringplanandthecoverageandeffectivenessofthewholeageingmanagementprogramme.

STUKreviewstheannualageingmanagementfollow‐upreportsofthelicensees.Report‐ing of the containment cable ageingmonitoring plan shall be presented at least everyfive years in connection of the follow‐up report. STUK inspectors make observationsabout thecableageingaspartof inspectionsduringoperationandoutagesof theNPPunits,inconnectionoftest,maintenanceandfaultrepairingworksofcablesandrelatingequipment.SofarthefindingsofSTUKobservationsofphysicalcablesandreviewsofre‐lating reports of faults,maintenance, repair and renewalworks of them have been incongruentwiththereportsoftheageingmanagementprogrammes.

In the early years of Loviisa power plant operation the cable agingmanagement pro‐cesseswerenotefficientenoughtoconfirmthepreservationofthecableconditioninallcasesand locations.Therewereunpredictedhotspotsexposingcableswithhighther‐malandradiationstresses.Theutilitytookthecorrectiveandpreventiveactionsasde‐scribedinsection3.2.

Tothestrongfeaturesofthelicensees’cableageingmanagementprogrammesbelongsmonitoring the temperatures in containment and using detectionmethods to find outthehotspots.Cablesampledepositsareputinrepresentativeplacesandsamplepiecesaretestedregularly.Testresultshaveledtothenecessarymeasuressuchasadditionalqualification or cable changes. In that way the ageing management processes for thesafetyrelatedcablesinsidethecontainmentsofLoviisaandOlkiluotoNPPsarecapabletomaintainthequalifiedstageandpredictthenecessarymeasuresforrequalificationorcablechanges.



Forthesafetyrelatedcablesoutsidethecontainmenttheperiodicaltestandinspectionprogrammes have turned out to be effective so that no significant degradation due toageinghasbeenreported.

For the reasons presented above STUK considers that the ageing management pro‐grammesconcerningthesafetyrelatedcablesoftheFinnishNPPshavebeenadequate.

4 Concealedpipework

There is actually notmuch of concealed pipework in the Finnish nuclear power plantunits. In the following the few cases of buried or embedded pipework are shortly ex‐plained:


CoolingwaterwaysinNPPunitsLoviisa1and2areeitherrockorconcretetunnels,sea‐waterpipingislocatedinconcretebuildingsorinconcretechannels.Concretechannelsareeitherincast‐in‐situstructuresormadeofprefabricatedelements.

TherearesomepipingofTG‐systemembeddedinconcreteinconnectionwiththepoolsof reactorbuilding.TG‐pipinghasbeenhandled inRI‐ISI agingmanagementprogramandthereisonlyafewmetersofpipingthatisembeddedinconcrete.


CoolingwaterwaysinNPPunitsOlkiluoto1and2aremainlyconcretechannels,whichhave theirownAMPs.Coolingwaterpiping is installedeither in concretebuildingsorchannelssothatitcanbeinspected.Therearefloordrainagepipesandsewagepipesinconcrete or buried underground inNPP units Olkiluoto 1 and 2. There are also somesafetyclassifiedpipes,whichhavepenetrationthroughthickconcretestructures.

Figure4.1showsacase, inwhicha structurehashadsmall leakageinacondensationwater pipe between the turbine and concrete channel to seaside. Efflorescence in thesurroundingconcretestructurehasbeenidentifiedintheageingmanagementprogramandthecorrespondingfloodingriskhasbeenevaluated.Furthermaintenanceworksarescheduledaccordingly.

Seawater cooling system inNPP unit Olkiluoto 3 ismainly using cast‐in‐situ concretechannels or piping installed in rock or concrete channels. Concrete seawater channelshave their own AMP FIN005‐CEC‐6350‐690132 “In service inspection plan of coolingwaterstructures”.



Figure4.1Caseoffloodingriskbecauseofefflorescence.

NPPunitOlkiluoto3

TherearesomesafetyclassifiedpipinginsideconcretestructuresinNPPunitOlkiluoto3.

IntheconcretebasematoftheOlkiluoto3containmentinsidethesteellinerthestainlesssteelprocesspipesofIRWSTtank(In‐containmentRefuellingWaterStorageTank)areinsideofa sleevepipe.Thesleevepipe formsapartof the liner. Inorder toallow themovementbetween the liner and the concretebasemat a stainless steel compensatorwas installed inside the liner. Outside the liner these straight IRWST process pipes(length7m)canbeinspectedinsideofthepipe.

TherearefloordrainagepipesandsewagepipesinconcreteorconcealedundergroundinNPPunitOlkiluoto3.Stainlesssteeldrainagepipesofthe fuelpoolsareinsidethickconcretestructures.

5 Reactorpressurevessels

5.1 DescriptionofageingmanagementprogrammesforRPVs

TheageingmanagementofNPPunitsLoviisa1and2RPVs isbasedonSpecialageingmanagementprogram[LO1‐K822‐00044version2.0].

TheageingmanagementofNPPunitsOlkiluoto1,2and3RPVsisdefinedinthegeneralageingmanagement programme for TVO’s nuclear facilities [OL1, OL2, OL3 and KPA:Document117279].TheprogrammedescribestheageingmanagementprocessfortheSSCs of TVO’sNPPunitsOlkiluoto 1, 2 and 3 and the spent fuel storage aswell. The



scopeofthisdocumentisquitewideandthereforeitisalsoquitegeneralincludingi.a.organizationalaspects.However,itgivesthegeneralageingmanagementprinciplesthatarefollowedatOlkiluotosite.

5.1.1 ScopeofageingmanagementforRPVs

ItiscommonknowledgethatachievingtheoptimalservicedesignlifeofaSCCrequiresthat it is operatedwithin its allowable operating limits. The ageing of nuclear powerplant SSCs that are important toplant safety andavailabilitymustbeeffectivelyman‐agedtoensurethattheirrespectivedesignfunctionsaremaintainedthroughouttheser‐vice life of the plant. This process involves the prediction anddetection of equipmentdegradationwhen required safetymargins are threatened, aswell as the initiation ofsubsequentcorrectiveormitigationactions.RPV, a SC1 component, forms a part of theRCPB and iswithin the scope of thisAMPevaluatedwith respect to ageingdegradationeffects and their consequences.All pres‐sure‐retainingandsurfaceweldsareincludedinthisreport.AgeingisaprocessbywhichthephysicalcharacteristicsofaSCCchangewithtimewhensubjectedtoaspecificenvironment.Ageingdegradationmayproceedthroughthepro‐gressionofasingleageingmechanism,oracombinationofseveralageingmechanisms.TheageingofmaterialsinanuclearpowerplantmayleadtofunctionaldegradationofplantequipmentincludingalsotheRPV.Thepressurevesselmaterialmustretainitsfractureresistanceevenwhenthematerialisaging.NPPunitsLoviisa1and2TheRPVshavebeenorderedwithoutanymaterialspecificationaccordingtothecurrentrequirements,soaccuratematerialinformationanditsimpactontheperformancechar‐acteristicsofthepressurevesselhasonlybeenevaluatedbyanalysesperformedlater.TheRPVsandassociatedprimarynozzlesaremadeof lowalloy high‐temperatureCr‐Mo‐Vsteel,15X2MΦA.Theoldmarkingforthissteelhasbeen12X2MΦA.Thematerialof theRPVcover ishigh‐temperaturesteel20X2MΦA.TheRPV flange ismadeofsteel25X2MΦA.Thepressurevesselismainlyweldedbysubmergedarcweldingusingfillermaterial C10XMΦT and powderAH‐42. Schematic drawings of LO1 and LO2RPV andcoverareshowninAnnex4.ThecompositionsofbasematerialsandweldingadditivesaresummarizedinTable5.1.

Table5.1.ChemicalcompositionofthemainLoviisaRPVsteelsandweldingfillermetal

C Si Mn S P Cr Ni Mo V As Co Cu

15X2MΦAvessel,nozzles

≥0.13≤0.18

≥0.17≤0.37

≥0.30≤0.60

≤0.025 ≤0.025 ≥2.50

≤3.3≤0.40 ≥0.60

≤0.80≥0.25≤0.35

≤0.08

≤0.025 ≤0.3



20X2MΦAtop

≥0.16≤0.21

≥0.17≤0.37

≥0.30≤0.60

≤0.025 ≤0.025 ≥2.50

≤3.0≤0.40 ≥0.60

≤0.80≥0.25≤0.35

≤0.08

≤0.025 ≤0.3

25X2MΦAtopflange

≥0.22≤0.27

≥0.17≤0.37

≥0.30≤0.60

≤0.025 ≤0.025 ≥2.80

≤3.30≤0.40 ≥0.60

≤0.80≥0.25≤0.35

≤0.08

≤0.025 ≤0.3

b10XMΦTweld

≤0.11

≥0.35≤0.45

≥0.55≤1.30

≤0.03 ≤0.03 ≥1.1

≤1.6‐ ≥0.40

≤0.60≥0.15≤0.30

‐ ‐ ‐

The pressure vessel consists of 7 cylindrical forgings. The RPV cover and bottom aremadeoftwopiecesthatareweldedtoeachotherbyanelectroslagweldingusingweld‐ingfillermaterialC13X2MTΦandpowder0Φ‐6.Topreventcorrosioncausedbyprimarywatertheinnersurfaceofthepressurevesseliscoatedwithanausteniticsteelcladding.Theinnersurfaceandpartlytheoutersurfaceofthecoverarealsocoated.Thefirstlayerisweldedbyapowdercoarsemethodusingtheso‐called"superimposed"additiveribbonC07X25H13andpowder0Φ‐10.ThesecondlayerwasmadebythesamemethodusingtheadditivetapeC07X19H10.Theweldingparametersofthesecondcoatinglayerarechosensothattheheatproduc‐tionnormalizesthefirstcoating layer, thereby improvingthepropertiesof thecoatinginterface.Thetotalthicknessofthecoatinglayeris8‐10mm.Thecoatingisweldedac‐cordingtoweldingstandardNPSOP‐1513‐70.ForLoviisa1thecopperconcentrationsoftheweldandbasematerialsare0.18%and0.15%andthephosphorusconcentrations0.030%and0.015%,respectively.ForLoviisa2thefiguresareslightlysmaller:0.10%and0.13%forcopperand0.025%and0.012%forphosphorus,respectively.Thereare17nozzlesinthecylindricalbodyoftheRPV.IntheRPVcover,thereare37penetrationsfortheCRDMsand18penetrationsforinstrumentation.TheCRDMnozzlesareweldedtothecoverfrominside.Sothereisacreviceopeningtooutsidebetweenthenozzle and the cover. The instrumentation nozzles arewelded to the outside coating.Theinstrumentationnozzlesaremadeofferritic(22K)forgings.TheCRDMnozzlesaremadeof pipe (CT20) and forged flange (22K). Inside thenozzles thereareprotectiveshieldpipesmadeofausteniticstainlesssteel.Thenozzlesofthemaincoolantlinesaremadeofforgingsthathavebeenmachinedtothe final shape.Themainparts are interconnectedby submergedarcweldingmethodaccordingtotheweldingstandardNPSOP‐1513‐70.NPPunitsOlkiluoto1and2TheRPViscomprisedofacylindricalshellwithaweldedbottomheadandaremovabletophead.Theshellandtopheadareconnectedwithflanges.Themainprocessandauxil‐iarypipenozzlesareweldedintothecylindricalshell.Thecontrolroddrivenozzlesand



maincirculationpumpnozzlesarelocatedinthebottomhead.TheRPVissupportedbyasupportskirtthat isweldedtothecylindricalshell.ThemaininternalsofBWRRPVsinclude:steamdryer,steamseparators,coreshroud,corespraypiping,fuelassemblies,control rods, control rod guide tubes and nozzles as well as the main recirculationpumps.TheRPVshavebeen fabricatedbyASEAAtom,which later becameABBAtomandpresentlyitisWestinghouse.TheheightofaBWRRPVisapproximately19600mm,theouterdiameter5660mm,andthewallthicknessisapproximately140mm.ThemostimportantchangesrelatingtotheworkingconditionsoftheRPVarethepowerupratings.Themostimportantpowerupratingwasimplementedin1997.TheoperationoftheRPVisdeliberatelycontinuedbeyondtheoriginaldesignlifetime.Inthiscontextalsothetransientsrelatingtothefatiguecalculationsarere‐evaluated.Both of the abovementioned changes increase the lifetime neutron fluence. This hasbeentakenintoconsiderationinthesurveillanceprogrammes.SupportstructuresweldeddirectlytoexternalsurfaceoftheRPVarenotinthescopeofthisAMP.The pressure‐retaining RPV subcomponentswithin the scope of this AMP include thelowerandupperhead,coreshell,flangeandnozzles.AschematicdrawingoftheRPVsoftheNPPunitsOlkiluoto1and2isshowninAnnex5andasummaryofitssubcomponents’materialsisshowninTable5.2.

Table5.2.ChemicalcompositionofthemainRPVsteelsandfillermaterialsatOlkiluoto1and2[Nevander,O].Reactorpressurevesselplate,ofASTMA533GradeB,Class1 C Si Mn P S Cr Ni Mo Cu Co SnMax. 0.20 0.30 1.50 0.015 0.02 0.30 0.70 0.57 0.15 0.05 0.05Min. ‐ 0.15 1.20 ‐ ‐ ‐ 0.45 0.47 ‐ ‐ ‐Flangeandnozzleforgings,materialequivalenttoASTMA508‐69,Class2 C Si Mn P S Cr Ni Mo Cu V Sn CoMax. 0.25 0.35 0.80 0.025 0.025 0.45 0.90 0.70 0.25 0.05 0.05 0.05Min. 0.18 0.15 0.50 ‐ ‐ 0.25 0.50 0.55 ‐ ‐ ‐ ‐Nozzlesinbottomend,ofamodifiedspecificationofmaterialtoSwedishStandardSIS2103 C Si Mn P S Cr Ni Cu Co N Max. 0.16 0.50 1.60 0.025 0.025 0.25 0.025 0.20 0.05 0.009 Min. ‐ 0.15 ‐ ‐ ‐ ‐ ‐ ‐ ‐ ‐ ThestainlesssteelcladdingattheinsidesurfaceoftheRPVcylindricalshellandthecov‐eris3mmthickatleast.Theweldtypeisof18/8includingnotmorethan0.06%Car‐bon,0.08%Nitrogenand0.05%Cobalt.



Thecladdingattheinsidesurfaceofthebottomis3mmthickatleastandthematerialisInconel182.NPPunitOlkiluoto3ThesupportsweldeddirectlytoexternalsurfacesoftheRPVarenotinthescope.The pressure‐retaining RPV subcomponentswithin the scope of this AMP include thelowerandupperhead,coreshell,transitionring,nozzles,flanges,andsafe‐ends.AschematicdrawingoftheOL3RPVisshowninAnnex6andasummaryofitssubcom‐ponents’materialsisshowninAnnex7.AMPevaluationboundariesfortheRPVextendtotheirconnectiontotheMainCoolantLinesandtheCRDMs.TheMCLsareconnectedtotheRPVattheinletandoutletnozzlesafe‐ends.TheevaluationboundaryforthisAMPreportendsattheRPVinletandoutletnozzlesafe‐endswheretheyareconnectedtotheMCLpiping.TheweldconnectingRPVinlet/outletnozzlesafe‐endandMCLpipingishandledwithintheMCLSAMP.The CRDMs are connected to the RPV at the CRDM Adaptor Flanges. The evaluationboundaryforthisAMPreportendsattheCRDMAdaptorFlangeswheretheyareconnectedtotheCRDMs.Only low‐alloy steel, stainless steel andNi‐base alloyshavebeenused tomanufacturetheRPVsubcomponents.Therefore,onlyageingmechanismswhicharerelevantforsuchmaterials are considered. Furthermore, only ageing mechanisms that could occur inprimarywater incombinationwithneutron fluencehave tobeevaluated. InTable5.3ageingmechanismsthatareexcludedinadvance,arelisted.Table5.3.Ageingmechanismsthatareexcludedinadvance

AgeingMechanisms SusceptibleMaterial ReasonforExclusion

CausticCorrosion CarbonSteel,Low‐AlloySteel

Notincontactwithprimarywater

Flow‐AssistedCorrosion CarbonSteel,Low‐AlloySteel


Denting All OnlyrelevantfortheSteamGeneratortubes

CausticStressCorrosionCracking

CarbonSteel,Low‐AlloySteel,AusteniticStainlessSteel

CausticSCCisnolongerarelevantageingmechanismforPWRs

OuterDiameterStressCorrosionCracking

All OnlyrelevantfortheSteamGeneratortubes

Strain‐InducedCorrosionCracking

CarbonSteel,Low‐AlloySteel




5.1.2 AgeingassessmentofRPVs

NPPunitsLoviisa1and2AgeingmanagementofLO1andLO2RPVsisbasedonSpecialageingmanagementpro‐gramme[LO1‐K822‐00044version2.0].TheoriginaldesignlifetimeofLoviisaRPVswas40yearsandnowitis50yearswhenthenewlicensingperiodhasbeengranted.ThemostimportantdegradationmechanismsandcriticalareasoftheRPVhavebeenidentifiedandlistedinthetable5.4.Identificationhasbeendonebylicensee'sexpertsorexpertpanelsandisbasedoninternalandexternaloperatingexperience.Astheirradia‐tionembrittlementofthecoreweldisthedominatingdegradationmechanismofRPVthecorporatelevelR&Dprogrammeshavefocusedremarkableeffortsonthisissue.

Table5.4CriticalareasanddegradationmechanismsofLoviisaRPV

RPVsection Criticalarea degradationmechanism

RPV‐head Penetrationnozzles(instrumentationandCRDMnozzles)Outersurface

FatigueFatigueofnozzlesleevewelds(effect:sleevedefor‐mation)SCCofweldsBoricAcidCorrosion

Mainflange Sealing surfacesThreadedholesformainstuds

BoricAcidcorrosionStresscorrosioncrackingFatigueofsealinggroovesFatigueCorrosionWearingofthreads

Vessel VesselbodyCladdingNozzles

Irradiationembrittlement(coreareaandweld)WearingofguidingsurfacesGrowthofexistingflawsFatigue

TheonlyimportantdegradationmechanismoftheRPVheadiscorrosioncausedbypri‐marywatercontainingboricacid.Thisphenomenon ispossibleonly in thecrevicebe‐



tweenthecoverandthenozzle.ThisispossibleonlyincasethereisaleakinpipesthatarelocatedabovethecoverlikeCRDMnozzles.LoosingleaktightnessofthesealingsurfacesoftheCRDMandinstrumentationflangesmay leadtocorrosion in ferriticmaterialof thenozzles. Theheadmayalsocorrode iftheleakislarge.SCCofausteniticmaterialcouldalsohappeninsomecases.Aprotectivecorrosionsleeveisweldedtothenozzleattheupperandlowerends.Duetodifferent thermal expansion coefficients there exist cyclic thermal stresses in the con‐struction.Thismayleadtolossofleaktightnessoftheweldsduetothermalfatigue.Thethreadedholesofmainflangeareingoodcondition.Thethreadsareloadedbynor‐maltighteningandalsobypotential impuritiesintheholes.It is importanttokeepthethreadsingoodconditionbecauseitisnoteasytorepairthem.Thetransitiontemperatureshiftduetoradiationandtheresultingchangeinthemateri‐alfracturetoughnessaremonitoredbyasurveillanceprogramme.Theradiationshiftisinfluencedbythecopperandphosphorusconcentrations.Thecriticalareaisthecorear‐eaweld4.As soon as the inspection programme was started, analyses of irradiation samplesshowedthattheembrittlementrateoftheRPVwashigherthanexpected.TheproblemiscrucialbecausethedesignoftheVVER‐440RPVisnarrowerthanmostRPVsinwesternreactors.Inaddition,adisadvantageous(highCu‐andP‐content)weldedseamislocat‐edinthecorearea.SincetheembrittlementoftheLoviisa1RPVwasabletoproceedtoolongbeforereduc‐tionofthecore,Loviisa1RPVhadtobeheattreatedin1996.Theradialguidesareweldedtothecladdingbyfilletwelds.Fatigueintheseweldsisanidentifiedmechanismrelating toageingmanagement.Therisk for fatigue isgreateratLO1RPVwithsharperweldscomparedtoLO2RPVwheretheweldshavebeengrindedsmoother.Anotheridentifiedageingmechanismiswearingoftheradialguidesurfacescausedbymovementsofinternalsduringoperationandoutages.Theintegrityofthecladdingespeciallyinthecoreareahasanimportantroleinprevent‐ingbrittlefracture.TheprimaryandECCSnozzlesarecomplicatedstructureswithbimetallicwelds.Fatigueis theirmost important identified failuremechanism.However, fatigue analyses showacceptablefatigueevenfor50yearsofservicelife.NPPunitsOlkiluoto1and2ThefatigueanalysesoftheRPVsofNPPunitsOlkiluoto1and2havebeenupdatedfor60yearsof service life.Atpresent, therearenoknownageingmechanisms that could



limit the technical service life of the RPVs after reaching the aforementioned plannedlifetime.The[IAEASafetyStandardsSeriesNo.NS‐G‐2.12]principleshavebeenusedinprepar‐ingtheagingmanagementprogramme.Thisprogrammeisevaluatedannuallyandup‐datedasnecessary.TheRPVswhohavelimitedlifetimewillneedtobecontrolledbyanalysesortests.Asanexample,loadanalysesaretobementioned,whichmustbeupdatedatthelatestinLTOphaseoriftheregulationsotherwisesorequire.BasedonoperationalexperiencethemostimportantageingmechanismatBWRplantsisstresscorrosioncracking(SCC).ThematerialInconel182andalsosomeoftheausteniticstainlesssteelsusedintheRPVrequiremonitoringofSCC.ThemostimportantsubcomponentsonSCCpointofviewarethenozzlesofsystems312(feed water system), 321 (shut‐down cooling system) and 323 (core spray system).Nowadaysthein‐serviceinspectionsaremadeat2yearintervals.ThematerialInconel182whichispronetoSCCisalsousede.g.inthelowernozzlesoftheRPV.However,theriskisherelowerbecauseofheattreatment.NPPunitOlkiluoto3The ageingmanagementofNPPunitOlkiluoto3RPV is basedon Special ageingman‐agementprogramme[FGFNTCM‐G/2009/en/1013rev.B].The fatigue analyses of the RPV is 60 years of service life. At present, there are noknown ageingmechanisms that could limit the technical service life of the RPV afterreachingtheaforementionedplannedlifetime.Table5.5containsasummaryoftheageingmechanismsthathavebeendeterminedtorequirefurtherevaluation.Table5.5alsoincludesalistofthesubcomponentsthatmaybeaffectedbythisageingmechanism.

Table5.5.Ageingmechanismsthatrequirefurtherevaluationforin‐scopesubcom‐ponents

RelevantAgeingMechanisms

SusceptibleMaterial AffectedSubcomponents

ThermalAgeing Ni‐BaseAlloy AllDMWs

Neutron‐InducedAgeing(IrradiationEmbrittle‐ment)

Low‐AlloySteelwithAusteniticStainlessSteelCladding

CoreBeltlineRegion(CoreShell)

AgeingUnderCyclicorTransientLoading(Fatigue)

AllMaterials AllRPVSubcomponents



BoricAcidCorrosion Low‐AlloySteel ExternalSurfaces

PittingCorrosion AusteniticStainlessSteelLow‐AlloySteel

AllAusteniticStainlessSteelSubcomponents,ExternalSurfaces

IntergranularCorrosion AusteniticStainlessSteel

VentPipe,VentPipeConnection,VentValves

CreviceCorrosion AusteniticStainlessSteelLow‐Alloy Steel

AllIn‐ScopeSubcomponents

IntergranularStressCorrosionCracking

AusteniticStainlessSteel

VentPipe,VentPipeConnection,VentValves,Coldworkedsub‐components(VentPipe)

PrimaryWaterStressCorrosionCracking

Ni‐baseAlloy Adaptortubes(bothCRDMandInstrumentation),Ventpipepenetration,Domethermocouplepenetration,DMWs

TransgranularStressCorrosionCracking

AusteniticStainlessSteel

AllAusteniticStainlessSteelSubcomponents

Environmentally‐AssistedFatigue

AllMaterials Allsubcomponentsincontactwithmedium

TheinformationinTable5.5isaddressedunder,whereeachageingmechanismunder‐goesevaluationtodeterminewhetheritisadequatelymanagedbyexistingplant‐specificAgeingManagementactivitiesandmeasuresduringthe60yearsserviceperiod.1.ThermalAgeingThermalAgeinghasbeenidentifiedasarelevantageingmechanismfortheOL3DMWsbetween:

radialguidetotransitionring inletandoutletnozzlesandthenozzlesafe‐ends adaptorstoupperheadandinstrumentation socketforventpipetoupperhead thermocouplepenetrationtoupperhead

TheinfluenceislimitedtotheDMWsbetween

theferriticRPVnozzlesandausteniticstainlesssteelsafe‐end componentsoflow‐alloybasematerialwithnickel‐basealloybuttering nickel‐basecomponents componentsoflow‐alloybasematerialwithausteniticstainlesssteelcladding nickel‐basealloycomponents austeniticstainlesssteelcomponents.



TheinitialassessmentofRPVDMWthermalageingsusceptibilityresultedintheconclu‐sionthattheonlyareathatcouldbesusceptibletothermalageingduringplantopera‐tionistheCarbon‐depletedzonelocatedalongtheHAZfusionlineofthelow‐alloysteel.ThiszoneisinducedbythediffusionofCarbonfromthelow‐alloysteelbasematerialtotheweldmetal, and formedduring stress relievingheat treatments performedduringcomponentassembly.Theadverseeffectonthemechanicalproperties isrelatedtothedurationofheattreatment.Thermal Ageing Surveillance Programme exists to evaluate the thermal ageing of theRPVDMWs.Theprogramme is targeted towardprovidingprojectedmaterialpropertyresults in the end‐of‐life condition, including the tearing resistance at 300°C and theequivalentRTNDTof theRPVnozzle tosafe‐endDMWs.Through thisprogramme, thethermalageingoftheDMWsbetweenthesafe‐endsandtheRPVnozzlesisevaluatedtopredictmaterialpropertyevolution.This Thermal Ageing Surveillance Programme of RPV Dissimilar Metal Welds imple‐mentedto identifydegradedDMWsduringplantoperation. Ifdegradation is identifiedfor the RPV nozzles and the nozzle safe‐ends DMW, targeted inspections or testingshouldbeperformedforallotherDMWs.With the performance of this Thermal Ageing Surveillance Programme, it can be con‐cluded thatThermalAgeing is adequatelymanaged for in‐scope components and sub‐componentsduringthe60yearsserviceperiod.2.Neutron‐InducedAgeing(IrradiationEmbrittlement)IrradiationEmbrittlementhasbeen identifiedasarelevantageingmechanismforOL3RPV.RPVsubcomponentssubjectedtothehighestamountofsignificantfastneutronex‐posure are the two core shells (item4 of inAnnex7) and other adjacent subcompo‐nents,suchastransitionring(item2inAnnex7),flange‐integratednozzleshell(item5inAnnex7),andthecorrespondingweldedjointsandausteniticcladding.Irradiationembrittlementisnegligibleforthesafe‐ends,CRDMs,CRDMinstrumentationadaptor nozzles, vent pipe nozzles, andNickel‐based alloys used for the radial guidesandweldingfillermaterial.Inthesecases,theend‐of‐lifefluenceistoolowtointroduceanychangeinmechanicalpropertiesormicrostructure.AnirradiationsurveillanceisperformedonOL3RPVsubcomponentswhichconstitutesthecorebeltlineregion.Directmeasurementsafterarepresentativeirradiationofferrit‐icbasematerialanddepositedweldmetalaredeterminedthroughtheuseofasurveil‐lanceprogrammeinordertoverifytheconservatismofthepredictedmaterialembrit‐tlement(throughcalculationofRTNDTvalues).WiththeperformanceofMonitoringof IrradiationEmbrittlement forCoreBeltlineRe‐gion, itcanbeconcludedthat IrradiationEmbrittlement isadequatelymanaged for in‐scopecomponentsandsubcomponentsduringthe60yearsserviceperiod.



3.AgeingUnderCyclicorTransientLoading(Fatigue)RPVsubcomponentsaresubjecttoAgeingunderCyclicorTransientLoading(Fatigue).In‐scopecomponentsandsubcomponentsweredesignedforadefinedsetoftransientconditions. ExistingThermal andMechanical Fatigue analyses are safety analyses thatusetime‐limitedassumptions,whicharesubjecttorevalidationwhentheoriginaldesignlife of a particular component is to be exceeded and/or real operational loads signifi‐cantlydifferfromspecifieddesignloads.FatigueMonitoring System instrumentation is installed on inlet and outlet nozzles in‐cludingthesafe‐end(i.e.,mixingzoneofcoldandhotwater)inordertocapturethefa‐tigue relevant thermal loads.The applicationof anappropriate fatiguemonitoring ap‐proachconstitutes theprerequisite to thedeterminationof the fatiguerelevantopera‐tionalthermalloads.Usingthiskindofinstrumentationandmonitoringofoperatingpa‐rameters(i.e.,temperature,pressure),itispossibletovalidatethatthedesignassump‐tionsaremetduringplantoperation,alsoallowingindicationthatloadtransientsnotan‐ticipatedduringthedesignstagehaveoccurredorrealloadsarelesssevereandlessfre‐quentasdesignloads.AccordingtoASMEXIinspections(VT,UT)areperformedduringISIonsusceptiblein‐scopesubcomponents.Thiswillalsohelptodetectindicationsduetofatiguedamage.Therefore,althoughit isnotexpectedthatfatigueof in‐scopeRPVsubcomponentswillbeafactorwithintheservicelifeoftheplant,thisageingmechanismhasbeenclassifiedasarelevantdegradationmechanismforRPVsubcomponentswithinthescopeofSAMP.Actionsshouldbetakentorevisitthisageingmechanismifadverseresultsareobtainedduringmonitoring,inspectionortestingactivities.4.BoricAcidCorrosion(BAC)AconsiderablelossofmaterialcausedbyBACwouldoccuronlyinthecaseofaccidentalleakageresultinginthelong‐termpresenceofconcentratedliquidBoricAcidatelevatedtemperaturesonexternalcarbonorlow‐alloysteelsurfacesinanenvironmentthatcon‐tainsoxygen.Relevantin‐scopeareasincludeexternalsurfacesoftheRPVcomposedoflow‐alloyferriticsteel(16MND5).Thelong‐termpresenceofBoricAcidonsuchsurfacesismitigatedforRPVsubcompo‐nentsthroughtheuseofaleakagedetectionsystemintheReactorBuilding.Annualvisualinspectionsarealsoperformedonexternalsurfacesofaccessiblecompo‐nents in the primary circuit and adjacent systemsprior to plant start‐up following anoutage,aimingtodetectanyBoricAciddeposits.ItcanbeconcludedthatBoricAcidCorrosionisadequatelymanagedforin‐scopecom‐ponentsandsubcomponentsduringthe60yearsserviceperiod,andnofurtheractionsarenecessary.



5.PittingCorrosionPitting Corrosion is relevant for low‐alloy steel with a passivating layer in high‐temperaturewaterwithincreaseddissolvedOxygen.Forausteniticstainlesssteel,Chlo‐ride‐inducedpittingisarelevantageingmechanism.Criticalconditionsdependonlocaltemperatures,thepHofthemedium,andtheconcentrationofChlorides.PittingcorrosionismitigatedintheprimarycircuitatOL3becausethemaincoolanthasasufficientlylowcorrosionpotential.Forinternalsurfacesmadeofausteniticstainlesssteel,thepresenceofhighChlorideconcentrationsismitigatedbyimplementationoftheOL3WaterChemistryMonitoring.Asdescribedinthe[ChemistryhandbookpartV],theChloride concentration during steady‐state normal operation is below 0.01 mg/kg,whichisfarbelowtheCl‐concentrationwherePittingislikelytooccur.CriticalconditionsthatcouldpromotePittingCorrosionofexternal low‐alloysteelsur‐facescouldexistifleakageweretooccurcontinuouslyoveralongperiodoftime.How‐ever,thevisibleeffectofthisincidentisthepresenceofBoricAciddeposits.WhenBoricAciddepositsarepresent,theleadingageingmechanismisBoricAcidCorrosion.Theac‐tivitiesormeasuresmanagetheoccurrenceofBoricAcidCorrosionduringnormaloper‐ation.Therefore,theoccurrenceofPittingCorrosioniscoveredinthisrespectbythesemeasuresaswell.A significant amountof PittingCorrosionof external austenitic stainless steel surfaceswouldonly takeplace in thepresenceof externalChloride sources.These sources aremitigatedatOL3bytheprohibitionofChloride‐containingtapes,markings,fluids,etc.So Pitting Corrosion is adequately managed for in‐scope subcomponents and compo‐nentsthroughimplementationoftheWaterChemistryMonitoring,theuseofChloride‐freechemicalsandmaterial.6.IntergranularCorrosionIntergranularCorrosionisarelevantageingmechanismfortheausteniticstainlesssteelventpipe,ventpipeconnectionandventvalves.OxidizingconditionsaremitigatedatOL3throughtheapplicationoftheWaterChemis‐try Programme. In addition, the application of Chloride sources is mitigated at OL3throughtheuseofchemicalsormaterialsthathavealowChloridecontent.So Intergranular Corrosion is adequatelymanaged for auxiliary RPV in‐scope compo‐nentsandsubcomponentsthroughimplementationoftheWaterChemistryMonitoringandtheuseofChloride‐freechemicalsandmaterials.7.CreviceCorrosionCreviceCorrosionhasbeenidentifiedasarelevantageingmechanismforRPV.CreviceCorrosionisnotanindependentcorrosionmechanism,buttheenhancementofcommoncorrosionmechanisms(e.g.,GeneralCorrosion,PittingCorrosion)inspecialcrevicecon‐ditions.



ExternalsurfacesoftheRPVmadeoflow‐alloysteelareonlysusceptibletoGeneralCorrosioninoutageconditionswhencomponentsorsubcomponentsaresubjecttothepresenceofmoisturein localatmosphericconditions. Increviceconditions,thisageingmechanismcanbeenhancedsothatitcouldalsoberelevantduringnormaloutageperi‐ods. However, in‐scope components and subcomponents are located in the ReactorBuilding,wheretheairiscontrolledintermsofhumidity,temperatureandpurity.CriticalconditionsthatcouldpromoteCreviceCorrosionofexternalRPVlow‐alloysteelsurfacescouldexistifleakageweretooccurcontinuouslyoveralongperiodoftime.However,thevisibleeffectofthisincidentisthepresenceofBoricAciddeposits.WhenBoricAciddepositsarepresent, theleadingageingmechanismisBoricAcidCorrosion.TheactivitiesormeasuresstatedearliermanagetheoccurrenceofBoricAcidCorrosionduringnormaloperation.Therefore, theoccurrenceofCreviceCorrosion is covered inthisrespectbythosemeasuresaswell.LeakagemonitoringisimplementedatOL3tomitigatetheoccurrenceoflong‐termleak‐age, serving to mitigate Crevice Corrosion on external surfaces. System Plant Walk‐downsofexternalsurfacesforallaccessiblecomponentsarealsoperformedeveryyearfollowinganoutagepriortoplantstart‐up.WithrespecttoCreviceCorrosionofinternalsurfaces,thepresenceofhighChlorideconcentrationsisgenerallymitigatedbyimplementationoftheWaterChemistryMoni‐toring. However, austenitic stainless steelmay become sensitive to Crevice Corrosionwhen impurities (e.g., Chlorides) accumulate in the crevice during operation. This canleadtoanincreaseinsusceptibilitytopittingorgeneralattack,whenthepHinthecrev‐icechanges.Aswaterchemistryismonitored,CreviceCorrosionisnotgenerallyofcon‐cernforin‐scopecomponentsandsubcomponents.SoCreviceCorrosionisadequatelymanagedforRPVin‐scopecomponentsandsubcom‐ponentsthroughimplementationoftheWaterChemistryMonitoring,LeakageMonitor‐ingandtheuseofChloride‐freechemicalsandmaterials.8.IntergranularStressCorrosionCracking(IGSCC)ThermalSensitizationIGSCC does not affect components in the PWR primary and secondary circuit duringnormaloperatingconditionsduetothelowdissolvedOxygencontentinthewater.ThewaterchemistryatOL3iscontrolledincertainsystemsduringnormaloperation,withalimitedOxygencontent.However,Oxygeningresscouldoccurduringoutageorstart‐upconditions.VisualExaminationVT‐1willbeperformedontheligamentsbetweencontrolrodpene‐trationsofclosurehead.WiththisVT‐1examination,discontinuitiesandimperfectionsonthesurfacesoftheauxiliarysubcomponentscanbedetected.Ifdegradationisidenti‐fiedonthesurface,targetedinspectionsortestingshouldbeperformedfortheauxiliarysubcomponents.Therefore,itcanbeconcludedthatVT‐1examinationontheligamentsbetweencontrolrodpenetrationsofclosureheadcoversmonitoringof:



Ventpipe, Ventpipeconnection,and Ventvalves.

Oncethescopeandtypeofdegradationhasbeenidentified,itshouldbedeterminedifperiodic inspections should be implementedwithin the ISI Plan tomonitor this issueduringthe60yearsserviceperiodinsusceptibleareas(i.e.,inthecasethatdegradationdue toageinghasoccurred).Furthermore,VT‐2andSystemPlantWalk‐downswillbeperformedaftereachoutage,sothatleakagewillbedetected.ColdWorkingIGSCCcanoccurin lowECPconditionsintheeventthatamaterialhasbeensubjecttoseverecoldworking. It ispossiblethattheventpipewithinthescopeof thisSAMPre‐portthatiscomposedofausteniticstainlesssteelhasbeencoldworked.Astheseverityofthiseffectisnotcurrentlyknown,theseareasrequirefurtherconsiderationormoni‐toringduringthe60yearsserviceperiod.Mostin‐scopecomponentsandsubcomponentsareperiodicallyinspectedthroughwalk‐downsaftereachoutagebeforeplantstart‐up,andarealsocontinuouslymonitoreddur‐ingoperationusingtheLeakageMonitoringSystem.So,IGSCC(duetocoldworking)isadequatelymanagedforin‐scopeausteniticstainlesssteelRPVsubcomponentsandcomponents,throughtheimplementationofactivitiesandmeasuresthatareinplaceatOL3suchastheWaterChemistryMonitoring,VT‐1,VT‐2,Leakage Monitoring System and System Plant Walk‐downs performed following eachoutage.9.PrimaryWaterStressCorrosionCracking(PWSCC)Along‐termdamageofPWSCCcannotbeexcludedonanyofthetodayacceptedmateri‐als in the caseof adverseoperation conditions.Therefore,PWSCC is a relevant ageingmechanismifasusceptiblematerialconditionexists(i.e.,chemicalcompositionandre‐sidualstresses)incontactwithPWRmaincoolant.Ultrasonictestswillbeperformedonthefollowingcomponents:

Dissimilarmetalweld(DMW)betweenradialguidetotransitionring DMWsbetweentheinletandoutletnozzlesandthenozzlesafe‐ends DMWsbetweenadaptorflangestosleeveweldandInstrumentation

TheultrasonictestontheweldsincludesalsothetestingoftheHAZ.TheweldsandtheHAZaremoresusceptibletodegradationthroughstress.Ifdegradationisidentifiedforthewelds, targeted inspectionsor testingshouldbeperformed forallotherweldsandforthetubes.Therefore,itcanbeconcludedthattheultrasonicinspectionontheweldsandtheirHAZalsocoversmonitoringofthecorrespondingtubes.



VisualExaminationVT‐1willbeperformedontheligamentsbetweencontrolrodpene‐trationsofclosurehead.WiththisVT‐1examination,discontinuitiesandimperfectionsonthesurfaceofcomponentscanbedetected.Ifdegradationisidentifiedonthesurface,targeted inspectionsor testing shouldbeperformed for theRPVpenetrations.There‐fore, it canbe concluded thatVT‐1examinationon the ligamentsbetweencontrol rodpenetrationsofclosureheadcoversmonitoringof:

Ventpipepenetration Domethermocouplepenetration Weldbetweenthermocouplepenetrationtodomethermocoupleguide DMWsbetweenadaptortubestoupperheadandInstrumentation DMWbetweensocketforventpipetoupperhead DMWbetweenventpipetoventpipeconnection DMWbetweenthermocouplepenetrationtoupperheadandtothermocouple

tubeSo,PWSCCisadequatelymanagedforin‐scopecomponentsandsubcomponentsthroughperformingUTandVT‐1examination,andimplementationofLeakageMonitoringSys‐tem.10.TransgranularStressCorrosionCracking(TGSCC)TGSCCisarelevantageingmechanismforausteniticstainlesssteelwithaNickelcontentbelow15%whenexposedtowaterabove50°CthatcontainscriticalamountsofChlo‐ridesanddissolvedOxygen.Water chemistry of the main coolant is carefully monitored by the Water ChemistryMonitoring System. During steady‐state normal operation, the concentration of dis‐solvedOxygeninthemaincoolantisbelow0.005ppmandtheChlorideconcentrationisbelow0.01ppm[ChemistryHandbookPartVMonitoring].Thesevaluesare farbelowthecriticalconcentrationsdiscussedearlier,andthereforeTGSCCofinternalsurfacesisnotofconcernduringthe60yearsserviceperiod.In‐scopeausteniticstainlesssteelcomponentsandsubcomponentsintheventpipethatare in contactwithmain coolant are periodically inspected throughwalk‐downs aftereachoutagebeforestart‐up,andcontinuouslymonitoredduringoperationbytheLeak‐ageMonitoringSystem.Therefore,TGSCCisadequatelymanagedfortheventpipe.TGSCConexternalsurfacesduetothepresenceofexternalChloridesources(e.g.,adhe‐sivetapes,lubricants),ismanagedbytheprohibitionofchemicalsandmaterialswhichcontaineitherdissolvablechlorideorchlorine,whichmightbereleasedaschloridedur‐ingdecomposition.With theperformanceof these activities (WaterChemistryMonitoring, LeakageMoni‐toring,walk‐downsandtheuseofChloride‐freechemicalsandmaterials),itcanbecon‐cluded thatTGSCC isadequatelymanaged forall in‐scopecomponentsandsubcompo‐nentscomposedofausteniticstainlesssteelandNickel‐basealloysduringthe60yearsserviceperiod.



11.Environmentally‐AssistedFatigueEnvironmentally‐Assisted Fatigue is a relevant ageing mechanism for all componentsandsubcomponentssubjectedtocyclicloadingorrepeatedtransientloadingundersim‐ultaneousenvironmentalimpact.Environmentaleffectscanresultininitiationandgrowthofcracks,whichmightleadtoleakageorevenrupture.Therefore,areduction intheFatigue lifebyenvironmentally‐assistedfatigueshallbeconsidered.Sofartherehavebeennodocumentedoccurrencesof fatigue failure inPWRsduetoenvironmentaleffects, thathavenotbeencoveredbygivenfatiguedesign.However,theU.S.NRCrequirestheconsiderationofenvironmentalfactorsforFatigueassessmentduringthedesignphaseofnewplantsandlong‐termop‐erationintheU.S.[USNRCRegulatoryGuide1.207],aswellasduringtheevaluationphasesupportingtheLong‐TermOperationforexistingplants[NUREG‐1801].Anyenvironmentaleffectsareconsideredduringthedesignphasethroughtheselectionof appropriatematerials,dimensioning, reduction instressesbydesign, furtherproce‐duralmeasures ormeasures during the fabrication phase (e.g. cladding, avoidance ofcrevices, etc.). Fatigue analysis results are available for in‐scope components and sub‐components.Allsensitivezonesbyfatigueanalysesareinferriticsteelswithcladding,exceptforthesensitivezonesintheCRDMadaptertubes.Soferriticsteelzonesarenotwettedbypri‐maryfluid,becausetheyareprotectedbythestainlesssteelcladding.Thus,theydonotneedtobeconsideredintheenvironmentaleffectassessments.However,experimentalevidenceindicatesthepresenceofanenvironmentaleffectontheFatiguelifeofausten‐iticstainlesssteelcladdingsaswell.The most sensitive zone for the in‐scope primary side are the Ni‐Cr‐Fe alloy CRDMadapter tubes,wetted by primary fluid. The cumulative usage factors (CUF) includingenvironmentaleffect(Fen‐factor)resultedinvalueslessthan1.00duringthe60yearsofservicelifeforin‐scopecomponentsandsubcomponents.Nevertheless,accordingtoASMEXIultrasonictestingandvisualinspectionsareconsid‐eredintheISIofsusceptiblein‐scopesubcomponents.Actionsshouldbetakentorevisitthisageingmechanismifadverseresultsareobtainedduringmonitoring,inspectionortestingactivities.So,itcanbeassumedthatEnvironmentally‐AssistedFatigueisadequatelymanagedforin‐scopecomponentsandsubcomponentsduringthe60yearsserviceperiod.

5.1.3 Monitoring,testing,samplingandinspectionactivitiesforRPVs

NPPunitsLoviisa1and2In accordancewith the requirements of [STUK – Guide YVL E.5], periodic inspectionprogrammeshavebeenpreparedonthebasisofASMECodeSectionXIregulations.The



inspectionmethods tobeusedcorrespond to the requirementsofASMEXIorFinnishregulationsonpressureequipment.The selection of the inspection technique and the verification of inspection systems(equipment,instructionsandpersonnel)arequalifiedoncase‐by‐casebasisandsubmit‐tedtoSTUKforapproval.Theinspectionperiodis10years,correspondingtotherequirementsofASMEXIpara‐graphIWB‐2400andFinnishpressureequipmentregulations.Inspected objects are divided into categories based onASMEXI paragraph IWB‐2500.Eachinspectionobjectissubjecttotherequirementsofitscategoryandapprovallimits.TheinspectionresultsareassessedaccordingtoASMEXIparagraphIWA‐3000(dimen‐sioningprinciples)andIWB‐3000(acceptance limits) if it isonlypossible forthecom‐ponent,structureandmaterial.Abovetheapprovallimits,theacceptabilityofthefaultissolvedonacase‐by‐casebasis.The impactandacceptabilityof the faultcanbedemonstratedbycompilingaseparatestrengthanalysisusinganalyticalmethodspresentedinASMEXI.Primarycircuit tightnessandpressure testsareperformed inaccordancewith regula‐tionsforFinnishpressureequipmentandtherequirementsofASMEXIparagraphsIWA‐5000andIWB‐5000.IRRADIATIONEMBRITTLEMENTNeutronradiationcausesembrittlementofferriticmaterials.Theembrittlementcanbenoticed as rising transition temperature anddecreasing of upper shelf toughness.Thesensitivityofthesteeltoradiationembrittlementdependsonsteelalloyingandimpuri‐ties,inLoviisaRPVstypicallyonCuandP.Embrittlementisafunctionofradiationdose.Forthisreason,embrittlementvariesbe‐tweendifferentlocationsintheRPVwall,dependingonthestrengthoftheflux.Inaddi‐tion,embrittlementchangesrapidlyinthewalloftheRPVasafunctionofthicknessandisstrongestatthesurfacebetweenthebasematerialandthecoating.Someofthestructuralchangesinmaterialstructurecausedbyradiationaredynamicallyrecovered by the reorganization of atoms in thematerial. Increasing temperature im‐provesthismaterialrecovery.Forthisreason,theamountofembrittlementdependsal‐soontheoperatingtemperatureoftheRPV.Heattreatment,whichtakesplaceattemperatures150°Cto200°Cabovetheoperat‐ingtemperature, isabletoeliminatemostoftheradiation‐inducedembrittlement.Duetotheheattreatment,thetoughnessofthematerialisrecovered.



TestsandresultsDuring themanufacturing of the RPV several non‐destructive tests have been carriedout.Inaddition,non‐destructivetestswereperformedalsofortestspecimensmanufac‐turedsimultaneouslyinthesameway.Non‐destructivetestsduringthemanufacturingoftheRPVincludeultrasonictestingandmagnetictestingforthebasematerial.Thecoatinghasbeensubjectedtodyepenetrantandultrasonicinspections.Theweldshavebeenx‐rayedtotheextentithasbeenpossible.Othercriticalweldshavebeensubjectedtoultrasonic testing.Thesurfaceofallweldshavebeen inspected.Theinspectionshavebeencarriedout inaccordancewithanapprovedqualitycontrolpro‐gramme.TheRPV is finallypressure testedafter the lastheat treatmentat apressurehigherthanthedesignpressure.Duringfabrication,baseandweldmaterialofthecoreareahasbeenreservedforbrittlefracturestudies.TestspecimensmadeofthismaterialareinstalledintheRPVinloca‐tionsofhighneutronflux.Radiationembrittlementismonitoredwiththeseirradiatedsurveillancespecimens.TheoriginalsurveillanceprogrammeThe embrittlement causedby irradiation in forgings,welds andheat affected zones ofweldsismonitoredbyasurveillanceprogramme.Theevaluationof embrittlement rate isbasedonCharpy‐V toughness testsperformedforbaseandweldmetalsandHAZbeforeandafterirradiation.EmbrittlementhasalsobeenstudiedbytensiletestingandCOD‐fracturetoughnesstesting.Basedonthefirsttestresults,changestothereactorcorewerealsomade.Someoftheoutsidefuelelementswerereplacedbysocalleddummyelementswhodonotcontainfuel.Basedonbrittlefracturestudiesthetemperatureofemergencycorecoolingwaterwasraisedandchangesweredoneinautomation.Theoriginalsurveillanceprogrammehasbeenexpandedandmodernisedbasedonexperience.Also irradiated specimens thathavebeenheat treatedbefore re‐irradiationhavebeenplaced in the sample chains. These samples have been used to investigate the re‐embrittlementrateoftheRPVbasematerialandwelds.

ThenewsurveillanceprogrammeforLoviisa1RPVafterheattreatmentAfter the heat treatment of the coreweld seam in 1996, new surveillance specimenswereinstalledintheLoviisa1RPV.Thissurveillanceprogrammeislimitedonlytomoni‐toringtheweldmaterial.Theprogrammeconsistsoftwothermalannealingsandthreeirradiationcycles.Thelengthofirradiationcyclesequalstothreeoperatingcycles,andtheentireprogrammelasts9years.ThematerialfortheprogrammehasbeenacquiredfromtheRPVmanufacturerand,basedontheexaminationsitisrepresentativefortheLO1coreareaweld.



BasedontheresultsofthenewsurveillanceprogrammeithasbeendemonstratedthatthemargintobrittlefactureofRPVweld4canbeconsideredacceptableevenafter50yearsofoperation.

NPPunitsOlkiluoto1and2System218(Irradiationtestspecimens)consistsoftestspecimencontainers,testspec‐imensandneutrondetectors.The test specimen containers consist of stainless steel pipeswhich, after having beenchargedwith test specimens, are filledwith inert gas (argon) and sealedwithweldedplugs.Thecontainersaredesignedinordertobesuspendedinthespecialchannelsbe‐longingtosystem212(Coresupportcomponents)inthereactordowncomer.Thetestspecimensareoftwotypes:

Charpy‐Vimpacttestspecimensforthedeterminationofthetemperatureatwhichthetransitionfromductiletobrittlefracturetakesplace,and

tensiletestspecimens(miniaturetestspecimens)forthedeterminationofthe

yieldstrength,tensilestrength,ruptureelongationandrupturecontraction.Duringtheirradiation,thetensiletestspecimensareprovidedwithcopperfillerstogivethesamelimitingvolumeastheimpacttestspecimens.Theneutrondetectorsareoffoiltypeandmeasuretheneutronflow.Thefoilsarecon‐tainedintightlyweldedcontainers.Eachtypeofreactormaterialisrepresentedbyfivesetsoftestspecimens.Eachsetcon‐sistsofapproximately21impacttestspecimensandfourtensiletestspecimens.Onesetisdesignedforreferenceandistestedinnon‐irradiatedconditionandonesetisusedforacceleratedtestinginatestpositionclosetotheoutsideofthemoderatortank.Thethreeremainingsetsareirradiatedintestpositionsclosetothereactorvesselwall.Adeep(10mm)axialindicationwasdetectedinOL2system312nozzle/safe‐endweldalreadyin2003orevenearlier.Sincethentheindicationhasbeeninspectedfrominsideeverysecondyearuntil2011.Nochangeofthedepthoftheindicationhasbeennoticedduring these years. The tests were comparable because they were always performedwiththesamequalifiedordinaryeddy‐currentandultrasonictestingmethodwithama‐nipulator.In2011thenozzlewasinspectedwithaphasedarrayUT,whichisamoresophisticatedtechnique.Basedonthistechniquethedepth15mmwasmeasuredforthesameindica‐tion.Thisnewfindingstartedaprocessthatfinallyleadtorepairingofthe312nozzle/safe‐endweld.OtherRPVnozzlesofsamekindwerealsoinspected.Basedonthesein‐spectionsitwasalsodecidedtorepaironenozzleweldinsystem323(seealsoSection5.1.4).



NPPunitOlkiluoto3Continued satisfaction of the basic safety requirements for OL3RPV operation duringthe60yeardesignlifeisassuredthroughtheapplicationofoperationalmonitoringac‐tivities,In‐ServiceInspectionsandadditionalsurveillanceprogrammes.A. MONITORINGMEASURESThe identificationof ageingmechanisms relevant for theRPV is done in Section5.1.2.Themonitoringprogrammescorrespondingtheseageingmechanismsarespecifiedbe‐low.Monitoringprogrammesconsiderthedirectorindirectrecordingofactualloadingsandoperationalparametersduringnormalplantoperation.Themonitoringofoperatingparameters covers the continuousmeasurementof relevantparameters, suchaspres‐sure, temperature, flow rates,water chemistry and levelmeasurements. Transients oftheseparameters,aswellasthefrequencyofoccurrenceofspecifiedloadingconditions,arealsorecorded.Monitoringsystemsareapartof thepreventiveageingstrategyemployed for theOL3NPPinordertomitigatethedegradationofphysicalandfunctionalpropertiesofSSCs.1. MonitoringofThermalAgeingofDissimilarMetalWelds(DMWs)ThermalAgeingisarelevantAgeingMechanismandisingeneraladegradationprocessthatcommonlycausesdeteriorationinmaterial’sstrength,hardness,ductilityortough‐ness.AsurveillanceprogrammeisrequiredfortheDMWsofOL3RPVto:

Verifythehypothesesmadeduringthedesignstageregardingthematerialprop‐ertyvaluesusedintheLeakBeforeBreakandFastFractureAnalyses

IdentifysafetymarginsallowingthereductionofIn‐ServiceInspection(ISI)

scopeorconsiderationoffutureplantlifeextensionTheprogrammeistargetedtowardprovidingprojectedmaterialpropertyresultsintheend‐of‐lifecondition,includingthetearingresistanceat300°CandtheequivalentRTNDToftheRPVnozzletosafe‐endDMWs.2. MonitoringofIrradiationEmbrittlementofCoreBeltlineRegionA further relevantAgeingMechanism is irradiation embrittlement.DuringRPVopera‐tion, the vesselwall (core beltline region) is subjected to neutron irradiation, causingmaterialembrittlement. This ageingmechanismis monitoredusinga surveillanceprogramme.



TheIrradiationSurveillanceProgrammeconsiders60yearsdesignlifeoftheOL3RPV,anddescribesthemethodologyusedtomonitortheeffectofin‐serviceradiationonthematerial properties of RPV components in the beltline region. The Surveillance Pro‐grammedeterminesthebeltlinecomponentmaterialpropertiesthroughmechan‐ical testsofbothnon‐irradiatedtestspecimens(initialstate)and irradiatedtestspeci‐mens(post‐irradiationstate).Specimensarelocatedclosertothecoreandareconsequentlyexposedtoahigherrateoffastneutronexposurecomparedtotheinsidewallofthereactorvessel.Therefore,themechanical properties of irradiated test specimens, and the related RTNDT shift, arerepresentativeofthevesselandallowaprojectionofitsRTNDTshiftlaterinthereactorvesselservicelife.ThesecapsulescanbewithdrawnafterthevesselheadandtheupperInternalshavebeenremoved.ThepurposeoftheRPVIrradiationSurveillanceProgrammeistoexperimentallyverifythetensileandfracturetoughnessoftheRPVshellmaterialsafterarepresentativeirra‐diationoftestspecimens(post‐irradiationtesting).3. MonitoringofPrimaryCoolantWaterChemistryPrimarycoolantwaterchemistrymonitoringisasupportingtooltowardidentifyingtheonset of corrosive ageingmechanisms in the RCS and the RPV aswell. Monitoring ofprimarycoolantchemistryisoneofthepreventiveageingmeasuresappliedtotheOL3NPP inorder tomitigate thedegradationof physical and functional properties of sys‐tems,structures,andcomponents.Themainobjectivesoftheprimarysidewaterchemistrycontrolare:

Minimizemetalreleaseratesandcorrosionproductconcentration, Optimizecorrosionproductmigrationandre‐deposition Limitthecorrosionrateoffuelcladdingmaterial Suppressdecompositionofwaterbyradiolysis Preventlocalizedcorrosion(SCC/pitting)throughthelimitationofimpurities

(chlorides,fluorides,sulphates)4. LeakageMonitoringWithrespecttotheRPV,aLeakageMonitoringSystemprovideshumidityandair tem‐peratureinstrumentation,installedontheRPVhead.Thefunctionoftheleakagedetection equipment is to detect deviations from normal operation and to providemeanstopreventtheprogressionofinitiatingeventstosituationswhereasafetyfunc‐tionisneeded.BetweentheReactorPressureVesselandclosureheadaspecific leakagemonitoringisimplemented.ThisjointisequippedwithtwoO‐ringstoserveasapartoftheRCSPres‐sureBoundary.AsealleakofflinedrainsfromthespacebetweenthesetwoO‐ringsdi‐rectlytotheVentandDrainSystem.IftheinnerO‐ringisleaking,atemperaturesensorwilldetectatemperatureincrease.



5. LoosePartsMonitoringALoosePartsMonitoringSystem(LPMS) is installed intheRPV,withthreemeasuringlocations at thebottom, fourmeasuring locationson the top and twomeasuring loca‐tionsontheheadoftheRPV.TheLPMS(JYF)isappliedtoprimarycircuitcomponentstoensuretheRCSintegritythroughtheearlydetectionoflooseandloosenedparts.6. VibrationMonitoringFurthermore, a Vibration Monitoring System (VMS) with four absolute displacementtransducersisinstalledontheRPVclosurehead.TheVMS(JYG)isinstalledforearlyde‐tectionofchangesinthevibrationbehaviouroftheRPVinternalsandcomponentstructures.7. FatigueMonitoringThefatiguebehaviouroftheprimaryinletandoutletnozzlesismonitoredusingappro‐priate tools.Themeasuringpoint ison theMCL loop3on thehot legnext to theRPVnozzle.Thebasic idea in automatic fatigue monitoring is the measurement andstorageoftemperatureandpressuredatawithrespecttotime.Onetargetistoidentifysuchphenomenaasthermalshockorthermalstratification.8. ManagementofBoricAcidCorrosionBoricAcidCorrosionisatypeofuniformcorrosionthatattacksprimarilycarbonsteelandlow‐alloysteelinthepresenceofhot,concentratedaqueoussolutionsofBoricAcid.Main coolant (i.e. boratedwater) leakage results in deposits ofwhiteBoricAcidcrystalsandthepresenceofmoisturethatcanbeobservedbyvisualinspection.Concen‐tratedBoricAcidintheformofasolutionorcrystalscouldbepresentwheremaincool‐antleakageissubjecttoevaporation.Exposureofcarbonsteelorlow‐alloysteelsurfacescouldresult incorrosionandadecrease in thestructural integrityof componentsandsubcomponentsthatformapartoftheReactorCoolantPressureBoundary.[IAEASafetyReportsSeriesNo.57]requiresplantprogrammesforidentificationofpre‐ventiveandmitigationactions.Timelycorrectiveormitigationactionsforextentofdeg‐radationarepossiblethroughmonitoringandtrending.Operatingexperienceshouldal‐sobeconsideredforageingmanagement.Accordingto[NUREG1801]itisnecessarytomonitortheconditionofthereactorcool‐antpressureboundaryforboratedwaterleakage.Therefore, it isassumedthataBoricAcidCorrosionmanagementwillbeimplementedatOL3.BoricAcidCorrosionmanage‐mentisanAgeingManagementactivitythatmonitorstheeffectsofBoricAcidCorrosionontheintendedfunctionofanaffectedstructureandcomponentbydetectionofboratedwaterleakage.TheReactorCoolantSystemcontainsboratedwaterthatcouldaffectexternalsurfacesofstructures and components in the primary circuit, as well as adjacent components.Therefore,theseactivitiescoveranystructuresorcomponentswithwhichboratedwa‐tercouldcomeintocontact.



ForthemanagementofBoricAcidCorrosion,bothleakagemonitoringandvisualinspec‐tions(systemplantwalk‐downs)shouldbeimplemented.9. MitigationofChloridesDuetothefactthatchlorideimpuritieshaveanimpactonmanyageingmecha‐nisms,chloridesshouldbemitigated.Therefore,inadditiontothemonitoringofcor‐rosionimpurityconcentrations,generalAgeingManagementactivitiesarealsoinplacetomitigatethecontaminationofmediumsandsteelsurfaceswithChlorides.Severaladditivesorchemicalsareintroducedtothemaincoolantduringnormalopera‐tion.Thehandlingandqualityof theseadditivesor chemicalsaregenerally controlledaccording to their respective requirements to ensure that no critical concentration ofharmfulsubstancescanenter theplantwhichcould leadto thedegradationofcompo‐nentsandsubcomponents.ThepresenceofexternalChloridesourcesispreventedbyprohibitionoftheuseofChlo‐ride‐containing tapes, markings, fluids, etc. Chemicals should be listed in database,whicharepermissibleornotpermissiblematerialsfordifferentareasofuse.10. ForeignMaterialExclusionFuelandequipmentfailurescanoccurduetoforeignmaterialintrusionintotheplantsystemsandcomponents.However,foreignmaterialisnotconsideredariskfortheRCPB.B. PERIODICPROGRAMMESFORINSPECTIONANDTESTINGTheperiodicinspectionprogrammesarealsoapartofthepreventiveageingmanage‐mentstrategy.1. SurveillanceProgramme[IAEA Safety Standards Series No. NS‐G‐2.6] requires that a surveillance pro‐gramme should provide assurance that the operational limits and conditions aremetduringoperation.Therefore,asurveillanceprogrammewillalsohelptodetecttrendsinageing.TheobjectiveoftheOL3RCSLsystemistoensurethefunctionalavailabilityofsystemsandcomponentsthatperformsafetyfunctions,withinoperationallimitsandconditions,topromptlydetectSSCdeterioration,aswellasanytrendthatcould leadtoanunsafecondition[NuclearPlantChemistryConference,PaperReferencen°167046].TheRCSLsystemincludestheperformanceofperiodicinspectionsandtestingtoestab‐lishanddirectmonitoring,andinspectionoftheRPVandallSSCsgenerally.For theageingmanagementassessmentofRPVsubcomponents themonitoringof cer‐tainplantparameters (e.g.water chemistry, temperature) and the special surveillance



programmeforthermalageingisevaluatedinthisSAMPreportwithrespecttosurveil‐lance.2. Pre‐ServiceInspectionsPre‐serviceInspections(PSI)areimportantindeterminingthebaselineconditionofcer‐tain in‐scope components, andwere performed during plant construction andwill beperformed during commissioning. These results, in combination with those obtainedduringtheconstructionandmanufacturingphases,formthebasisforcomparisonwithsubsequentIn‐ServiceInspectionresults.3. In‐ServiceInspectionsIn‐Service Inspections (ISI) are performed forNPP components to provide continuingassurancethattheywillsafelyoperatethroughoutthelifetimeoftheplant,contributingtoplantavailabilityandminimizingtheprobabilitythataneventwilloccur.ISIactivitiesallow the comparison of baseline information to assess changes or degradation inequipmentconditionwhichhaveoccurredwithtime.Thecomponentselectioncriteriahaveevolvedtoincludeinspectionsthatshouldbeper‐formedatlocationssusceptibletodevelopingin‐servicedefects,andshouldbeimple‐mentedbasedon[STUK–GuideYVL3.8].

OperatingExperiencefeedback(bothplant‐specificandindustrywide) PotentialFailureMechanisms Risk‐informedarguments OperatingConditionsandStressanalyses(e.g.stressesorhighusagefac‐

tors,geometricaldefects) EquipmentDesign

The specific PSI / ISI requirements and ISI programme for theMain Primary Compo‐nentsaredefinedinrelevantinspectioninstructions.TheRPVISIPlanisbasedonASMEXIinspectionprogrammeB,IWA2432[ASMEBoiler&PressureVesselCode,SectionXI].Theintervalsforevenlydistributedinspectionsmaybechosenfromafewyearstoabouttenyears.Intervals’lengthshouldbechosenonthebasisofconservativeassumptions,toensurethatanydeteriorationofthemostexposedcomponent isdetectedbeforeitcanleadtofailure[IAEASafetyStandardsSeriesNo.NS‐G‐2.6].Generalproceduresor techniquesappliedduring ISI activitiesare reportedwithin theISIsummarydocument.OtherrelevanttestingproceduresforISIareperformedasfol‐lows:

VolumetricExaminations:UltrasonicTesting(UT)isusedtodetectvolumetricdefectswithinmetallicma‐terials(ASMEXIIWA2230).



SurfaceExaminations:‐MagneticParticleTesting(MT)isusedtodetectsurfacedefectsinmagnet‐icmaterials(ASMEXIIWA2220)‐LiquidPenetrantTesting(PT)isusedtodetectsurfacedefects(e.g.cracks)(ASMEXIIWA2220).

VisualExaminations:

Visualexaminationisperformedtogetinformationaboutthegeneralconditionofthepart,componentorsurface.Defectslikescratches,wear,cracks,corrosionorerosiononthesurface,orevidenceofleakageareobservable.

EddyCurrentTesting:

EddyCurrentExamination(ET)isusedtodetectvolumetricandsurfacedefects(ASMEXIIWA2233).

5.1.4 PreventiveandremedialactionsforRPVs

NPPunitsLoviisa1and2Themost important degradationmechanism of Loviisa RPVs is irradiation embrittle‐mentof theRPVcoreareaand itsweld.Therefore themost importantpreventiveandremedialactionsarerelated to that.Already thevery firstsurveillanceprogrammere‐sultsshowedthattheembrittlementrateoftheLoviisaRPVswasmuchfasterthanex‐pected. Therefore immediate mitigation actions had to be implemented in the earlyyearsofoperation.TomitigatetheexcessiveembrittlementoftheLoviisaRPVsfollowingactionshavebeenimplemented: designmodificationsoftheactivecore

o corereductionbyreplacingtheoutmostfuelelementswithdummyelementsofstainlesssteel,fluxreduction83%

o introducinganewfuelloadingpatterntoachievealowleakagecoredesign,fluxreduction18%

ThefollowingactionsandplantmodificationshavebeenimplementedinordertoreducetheriskofbrittlefractureoftheRPVduringaccidentortransientconditions: temperatureincreaseofemergencycorecoolingwater decreaseofheadandflowrateofhighpressureinjectionpumps increaseofpressurizerreliefvalvecapacity introductionofasignalfromhighprimarysystempressuretostopthehighpressure

pumps modificationofprotectionsignalsassociatedwithoccurrenceofpotentialsteamline

break thermalannealingofLoviisa1RPVcoreareaweld



FatigueisanotherdominatingdegradationmechanismoftheRPVs.RPV‐headnozzlere‐pairshavebeenperformedbecauseoffatiguedamagesand/ormanufacturingdefectsinthenozzlesleevesofbothunits.RPVmainflangesealingsurfacesofbothunitshavebeenrepaired.Mainreasonfortheserepairshavebeenlow‐cyclefatigueofthesealinggroovescombinedwithmanufacturingdefectsandcorrosioneffects.Thesefailureshavebeendetectedinperiodicinspections.NPPunitsOlkiluoto1and2Thepreventiveactionsinclude: assessmentofsusceptibilitytodegradationmechanisms, assessmentofpropagationofdegradationmechanisms,and protectiveactions.Theassessmentofsusceptibilitytodegradationmechanismsismainlybasedonmaterialtype, local geometry, prevailing stresses or strains and process chemistry. Recently, athoroughstudy[Cronvall,O]onthisissuehasbeendonefortheRPVsandtheirinternalsofNPPunitsOlkiluoto1and2,seeAnnex5fortheconstructionandsubcomponents.Forthesusceptiblelocations,degradationpropagationanalyseshavebeendone,cover‐ingallsignificantdegradationmechanisms.Theseincludeirradiationembrittlement,fa‐tigue, SCC and irradiation accelerated SCC (IASCC). These analyses are presented in[Cronvall,O]andtheycoveralsotheLTOperiod,i.e.continuedoperationfrom40to60years.Accordingtotheconservativeanalysisresults,thedegradationintermsofcrackgrowthisinmostcasesveryslow.Protectiveactionsconcernmainlyshieldingcomponents.IncaseofRPV,themostsignif‐icantshieldisthefeedwatersparger.ItdistributesthefeedwatermoreevenlytotheRPVdowncomerandalsoactsasa thermal shield to the feedwaternozzle andconnectingcomponents.Theremedialactionsinclude:

repair waterchemistrycontrol surfacetreatment componentreplacement

Repairingconcernsmainlyrepairwelding.Forinstance,in2017localrepairweldshavebeendonetoonesystem312nozzle/safe‐endweldandonesystem323nozzle/safe‐endweld.ThepipenozzleswererepairedbyamethodthatwasnotpreviouslyusedinFin‐land.TVOinitiatedtherepairasplannedbycorrectingthedetectedflawsinOL2nozzles.Therepairwascarriedoutbymachiningtheinsideweldandweldinganewfillercoatingwithafillermateriallesssusceptibletostresscorrosioncracking.



Water chemistry control concerns improving the chemistry andoxygen content of theprimary circuitwater todecrease the effect of degradationmechanisms,mainly to re‐duceorpreventinitiationandpropagationofSCC.Surfacetreatment,suchaslaserpeeningandshotpeening,introduceacompressivesur‐faceresidualstress.ThiseffectivelypreventstheonsetofbothSCCandfatigue.As forcomponentreplacement, itobviouslydoesnotconcerntheRPV.However,someinternalsarereplacedperiodically,e.g.controlrodsandfuel.AlsotheRPVmainflangeboltsareregularlyinspectedandreplacedifnecessary.

NPPunitOlkiluoto3Maintenanceisdefinedasallactivitiesnecessarytoestablishandkeeptheplantintheconditionrequiredtofulfilitsdesignatedfunctions[IAEASafetyStandardsSeriesNo.NS‐G‐2.6].Therangeofmaintenanceactivitiesincludesservice,overhaul,repairandre‐placementofparts,andoften,asappropriate,testing,calibrationandinspection.Therelevantmaintenanceactivitiesmaybedividedinto[IAEASafetyStandardsSeriesNo.NS‐G‐2.6]:

Preventive(includingperiodic,predictiveandplannedmaintenance),and Correctivemaintenance.

Thetaskofmaintenanceistopreventandtocorrectveryearlytracesofwearout,mate‐rialweaknessesorother typesofdegradationwhichmighthave anegative impactonsafetyoravailabilityoftheRPV.MaintenanceofcertainSSCswillbeperformedtoensurethattheconditionandfunc‐tionalityoftheSSCremainwithinacceptablelimits.Therefore,TVOwillapplyaspecificpreventiveconditionassessment strategyfortheOL3RPVthatinvolvesthefollowingactivities:

Recognizedegradationprocessesinatimelymanner(i.e.promptly),and Takemeasuresintimetopreventanegativeinfluenceonnuclearsafety,orto

sufficientlyminimizesuchinfluence. Shouldthemonitoring,testingorinspectionsshowresultsthattheRPVmay

havesufferedfromunexpecteddegradation,preventiveorremedialactionsmustbeconsidered.Theactionsshouldbebasedonarootcauseanalysiswhichin‐cludesthepossibledegradationmechanisms,theirorigin,evolutionscenarioandimpacttotheplantsafetyandavailability.ThepreventiveorremedialactionsarethenevaluatedandimplementedusingprocessesdescribedintheTVO’sop‐erativeinstructions.



5.2 Licensee’sexperienceoftheapplicationofAMPsforRPVs

NPPunitsLoviisa1and2Management of the embrittlement rate of the RPV core areaweld has been themostchallengingageingmanagementissue.CorrectiveactionshavebeentakenandthemostimportantonehasbeenthermalannealingoftheLoviisa1RPVcoreareaweld.TheheattreatmentofthecoreareaweldoftheLoviisa1RPVrestoredthetoughnessoftheweldmaterial close to the original level. The heat treatmentwas carried out at atemperaturerangeof475‐500°C. Insafetyanalysesavalueof 30°C isappliedto thepost‐heat residualbrittleness.The re‐embrittlementof theweldmaterial ismonitoredbyanewsurveillanceprogrammethatwasstartedin1996.Asaresultofextensivecorporatelevelresearchworkrelatedtotheirradiationembrit‐tlementthelicenseehasobtainedvaluableknow‐howandcompetenceinthisarea.ThishashadanimportantroleinsuccessfulageingmanagementoftheRPV.SeverallinearindicationshavebeenfoundinthesealingfacesoftheRPVmainflangein2005and2006.Theindicationswereopenedandfilled.AMPsfortheRPVshavebeenmodifiedslightlyduringplantoperationandaccordingtocurrent understanding they are comprehensive enough and adequate to manage theidentifiedageingmechanismsoftheRPVs.OperatingexperiencehascausedchangestotheinspectionprogrammesoftheRPVs.In theearly2000’sRPV‐headnozzle failureswerediscovered inoneof the twoVVER‐440reactorsandbasedonthisexperienceLoviisainspectionprogrammewasupdated.LaterinspectionsshowedotherfailuresofsimilarkindinLoviisareactors.However,af‐terthatnofurtherfailuresofthatkindhavebeendetected.In 2010 four indications close to each other were found in the circumferential weld(weld8)inLoviisa2RPVcover.ThelocationoftheweldisshowninAnnex4.Thecom‐bined size of these indications exceededASMEXI acceptance criteria. The approval oftheindicationswasshownlaterwithfracturemechanicalanalysis.Itwasassumedthattheindicationsweremanufacturingdefectsandthusnotpronetogrow.However,theirpotentialgrowthismonitoredregularlywithISI.In2016one linearaxial indicationwas found ina lowpressuresafety injectionnozzle(socalledTH‐nozzle)ofLoviisa1RPV.ThenozzleislocatedbetweenthecoldlegnozzlesofMCLs.Theindicationislocatedininsidecornerofthenozzlein321Olocation.Thein‐dicationwassizedtobe14mm(depth)and44mm(length).Afracturemechanicalanal‐ysisaccordingtoASMEXIwasperformedfortheindicationandbasedonthisanalysistheindicationcouldbeconsideredacceptable.Theindicationwasre‐inspectedin2017and the results did not show any growth.Due to development of the inspection tech‐niquetheindicationwasre‐sizedmoreaccuratelyandthenewsizeoftheindicationwassmaller,10mm(depth)and30mm(length).Theindicationwasreportedasindicationlocatedincladdingwhichmostprobablymeansseveralsmalloldweldingdefectsinsidethecladding. TheutilityhasinspectedallnozzlesofsamekindatbothRPVs.Sofarno



corresponding indications have been found. The utility is developing the inspectiontechniquefurtherinordertoimproveitsaccuracyandreliability.NPPunitsOlkiluoto1and2ThebasicevaluationandtechnicalbasisoftheAgeingManagementProgrammes(AMPs)according to IAEA is described in the table below. Additionally a short explanation isprovidedonhowtheissuewashandledin[Cronvall,O.SusceptibilityofBWRRPV].

Definition Handledin[Cronvall,O.SusceptibilityofBWRRPV] 1. Scopeoftheageingman‐

agementprogrammebasedonunderstandingageing

Allpossibledegradationmechanismsaswellastheassociatedcomponentspecificsusceptibility.Fa‐tigueisnotanissueintheRPVexceptforthefeed‐waternozzle.Thefeedwaternozzleishandledinaseparatereport

2. Preventiveactionstomini‐mizeandcontrolageingdegradation

Effectofimprovingwaterchemistryisdescribed,notmuchelseiscoveredconcerningthisissue.Nonewimprovementmethodshavebeenrecognized.

3. Detectionofageingeffects Inspectionsandappliedinspectiontechniquesaredescribedindetail

4. Monitoringandtrendingofageingeffects

Computational modellingofsignificantdegradationmechanismsisdescribedindetail,thecomputa‐tionalpartcoversthemallforallsignificantinter‐nalsandtheRPV.Thefeedwaternozzlefatigueishandledin[OL1andOL2‐TURO2018‐System312]

5. Mitigatingageingeffects Applicablemitigationtechniquesaredescribedindetail

6. Acceptancecriteria ThisissueisdiscussedwhenweighingtheTLAAresultsforallsignificantinternals,theRPVandthefeedwaternozzle.

7. Correctiveactions Thisissueisdiscussedonlybriefly.8. Operatingexperiencefeed‐

backandfeedbackofre‐searchanddevelopmentresults

Thisissueisnotincluded in[Cronvall,O.Suscepti‐bilityofBWRRPV].SomedamageshaveoccurredinthehistoryoftheRPVandinternals.Crackshavebeenobservedinsomenozzles,consolesandliftinglugs.Thesehaveallbeenhandledatthetimeanddocumentedindedicatedreports.

9. Qualitymanagement Thisissueisnotincluded in[Cronvall,O.Suscepti‐bilityofBWRRPV]asitisnotpartofthescope.

AgeingManagementProgramme[IAEASafetyReportsSeriesNo.82]fortheOL1andOL2RPV,internalsandthemainpipingsystemsarelistedbelow:

1. AMP101FatigueMonitoring:

ForallmainpipingsystemsinthecontainmentafullreanalysiswasmadeandallassociatedreportsweresubmittedtoSTUKaspartofthelicenserenewalproject.



Necessaryactionsfollowingtheseanalyseshavebeenlistedandwillbecarriedoutinthenearfuture.Apartfromthefeedwaternozzle[OL1andOL2‐TURO2018‐System312]thefatigueusageoftheRPVisinsignificant,nospecificAMPisneces‐sary.

a. Thepurposeofthisprogrammeistomanagelow‐cyclefatigueconsideredinthenuclearplantdesignbasis.Thisprogrammeappliestopipingcom‐ponentssubjecttocyclicloading.Thisprogrammemanagesalsoanyothertransientsduetotheoccurrenceoffluidconditionshavingthepotentialforinducingcyclicthermalstresseslike:

i. Stratifiedflow:accordingtothelatestCFDanalysesnostratifica‐tionoccurs,eveninthelasthorizontalpartofthefeedwaterpipebeforetheRPV.

ii. Swirlpenetration:nonsuchisrecognizedinthepipingorintheRPV.

iii. Thermalmixing:thermalmixingismorerelatedtohighcyclefa‐tigueandishandledinpurposemadereports.NonsuchispossibleintheRPV.

2. AMP102In‐serviceInspection/PeriodicInspection:FullASMEISIisperformed.Incaseoffindingstheprogrammeisintensified.Thisprogrammegenerallyincludesperiodicvisual,surface,and/orvolumetricexami‐nationandleakagetestofallClass1,2,and3pressure‐retainingcomponentsandtheirintegralattachments.Repair/replacementactivitiesforthesecomponentsandacceptancebyanalysis(suchasflawanalysis)arealsocovered.Thepro‐grammehasproventobeeffectiveinageingmanagementinClass1,2or3pipingandcomponents.

3. AMP103WaterChemistry:Themainobjectiveofthisprogrammeistomitigatelossofmaterialduetocorro‐sion.Thisincludesflow‐acceleratedcorrosion(FAC),stresscorrosioncracking(SCC)andrelateddegradationmechanisms,andreductionofheattransferduetofoulingincomponentsexposedtoatreatedwaterenvironment.Theprogrammeincludesperiodicmonitoringofthetreatedwaterandcontrolofknowndetri‐mentalcontaminantsbelowthelevelsknowntoresultinlossofmaterialorcrack‐ing.Thewaterchemistryprogrammeforalltypesofnuclearpowerplantswithwater‐cooledreactorsreliesonmonitoringandcontrolofreactorwaterchemistrybasedonseveralguidelines,suchasIAEASafetyGuideSSG‐13.

4. AMP104ReactorHeadClosureStudBolting:RegularISIandexchangeincaseoffindings.NospecificAMP.

5. AMP105BWRVesselIDAttachmentWelds:RegularISIandanalysisincaseoffindings.NospecificAMP.Thisprogrammein‐cludes

a. Inspections,inspectionrecommendationsandflawevaluationtoprovidereasonableassuranceofthelong‐termintegrityandsafeoperation.

b. EvaluationmethodologiesfortheattachmentweldsbetweenthevesselwallandvesselIDbracketsthatattachsafety‐relatedcomponentstothe



vessel.

6. AMP106BWRFeedwaterNozzle:FullASMEISI,intensifiedduetofindings.In2017repairofinnersurfacebreakingaxialIGSCCcrack.Thisprogrammeincludesenhancedin‐serviceinspectionslikeperiodicultrasonicinspectionsofcriticalregionsoftheBWRfeed‐waternozzle.

7. AMP107BWRStressCorrosionCrackinginCoolantPressureBoundaryCompo‐nents:ThemainobjectiveofthisprogrammeistomanageSCC,particularlyIGSCC,inBWRcoolantpressureboundarycomponentsmadeofstainlesssteel(SS)andnickel‐basedalloys,includingwelds.ProgrammesandgoodpracticestomanageSCCinBWRsaredelineatedinvariousnationalandinternationalreports,suchasIAEATechnicalReportNo.NP‐T‐3.13.FullASMEISI,intensifiedduetofindings.Regularpre‐analysisisperformedforpostulatedIGSCCcracks.Theprogrammeincludes:

a. preventivemeasurestomitigateIGSCCb. inspectionc. flawevaluationtomonitorIGSCCanditseffects.

8. AMP109BWRVesselInternals:

Thisprogrammeincludesinspectionandflawevaluationsinconformancewiththeapplicablerequirementsorguidancedocuments.Itaimstoprovidereasonableas‐suranceofthelong‐termintegrityandsafeoperationofBWRvesselinternals.Inaddition,thisprogrammeaddressesageingmanagementofBWRvesselinternalsofcastausteniticstainlesssteels(CASSs).Thisprogrammeconsiderslossoffrac‐turetoughnessduetoneutronembrittlementorthermalageingembrittlement.ThisprogrammealsoaddressesageingdegradationofX‐750alloyandprecipita‐tion‐hardened(PH)martensiticSSmaterials,andthoseBWRvesselinternalsthatareofmartensiticSSs.

9. AMP138ReactorCoolantPump:Allpumpswillbeexchanged2016–2019.Normalinspectionandmaintenance,nofurtherAMP.Thisprogrammeonageingmanagementcoversseveraldegradationmechanismsitmaybesubjectedtoandtheactivitiesnecessarytomanagetheage‐ingmechanisms.Assuch,thisprogrammereferstootherdegradation‐specificand/ormonitoringtypeofprogrammesthatdealwithparticulardegradationmechanismsanddegradationageingeffects.Thebodyofthepumpincludingitssealingpartsissafetyclass1,theinternalsofthepumpsperformingtheactivefunctionofthepumparesafetyclass2,andtheyareincludedinthescopeforLTOinaccordancewiththeIAEASafetyReportNo.57.

NPPunitOlkiluoto3TheAMPsforOL3areunderpreparation.In2017,OL3isinthecommissioningphaseandtheprimarycircuithydrostatictest(233barg)hasbeenperformedsuccessfullyinJune28,2017.Inaddition,theloadingconditionmonitoringprogrammehasstartedandwillcontinueuntilPTOconductedbythesupplierandafterPTOthemonitoringwillbecarriedoutbyTVO.



5.3 Regulator’sassessmentandconclusionsonageingmanagementofRPVs

RequirementsrelatingtoageingmanagementarebasedonSection5ofSTUKRegulationontheSafetyofaNuclearPowerPlant(STUKY/1/2016):

Thedesign,construction,operation,conditionmonitoringandmaintenanceofanuclearpowerplantshallprovide fortheageingofsystems,structuresandcomponents important to safety inorder toensure that theymeet thedesign‐basisrequirementswiththenecessarysafetymarginsthroughouttheservicelifeofthefacility.Systematicproceduresshallbeinplaceforprevent‐ingtheageingofsystems,structuresandcomponentswhichmaydeterioratetheir availability, and for the early detection of the need for their repair,modificationandreplacement.Safetyrequirementsandapplicabilityofnewtechnologyshallbeperiodicallyassessed,inordertoensurethatthetechnol‐ogyappliedisuptodate,andtheavailabilityofthesparepartsandthesys‐temsupportshallbemonitored.

Moredetailedrequirementsaregivenin[STUK–GuideYVLE.4]:

204.Forplantconstructionandmodificationprojects,thisGuidepresentsastrengthanalysiscoverageand reliabilityverificationprocedureconsistingof pre‐operational tests andmeasurements.Monitoring during service lifecovers themaintenanceof recurring loadswithin fatigueanalysisassump‐tions,andtheeffectsoftheoperatingenvironmentonthemechanicalprop‐ertiesofmaterials,specialemphasisbeingonradiationembrittlementofthereactorpressurevessel.

404.Obligations relating to theway of reporting the plans and results ofloadmonitoringaredeterminedonthebasisofSTUKrequirementsforpre‐operationaltestingandageingmanagement.

426. It shall be possible to compare the observations made during pre‐operational testingwith the results of loadmonitoring conducted duringservice. Supplementarymeasurement data on the behaviour of equipmentassemblies and significant local stress conditions shall be obtained,whichsupport thespecificationofthese loadingsbasedonthemeasurementdatamonitoredduringservice.

427.Operationalconditionsandevents inducing fatigue loadson themostimportantpressureequipmentshallberecordedover theservice lifeof thenuclearpowerplant.Sufficientmeasurementdatashallbecollectedontheirprogressforlaterverificationofessentialfactors.Themonitoringshallbesoarrangedthatunexpectedloadtypesoccurringduringoperationarealsode‐tected.

816. In themonitoringof loadandageingeffects, thequalitymanagementprocedures determined for ageing management in nuclear power plantsshallbefollowed.Thepersonnelevaluatingtheresultsshallhavegoodexpertknowledgeandexperience in the fatigueanalysesaccepted for thenuclear



powerplantinquestionaswellasitsbehaviourduringdifferentoperationalandtransientconditions.

903.STUKoverseesthemonitoringoffatigueloadsandvibrationsbyinspec‐tionvisitsduringpre‐operationaltestingandoperation,andalsobyreview‐ingannualreports.

STUKhasinspectedtheageingmanagementprogrammesincludingRPVageingmonitor‐ing plans. Implementation of the ageing management programmes has been verifiedduringSTUKinspections.

The ageingmanagement processes described in the ageingmanagement programmeswerefoundsatisfactoryattheoperatingNPPs.

TheageingmanagementprogrammesofOlkiluoto3arepracticallyfinishedalthoughtheplantunitisstillundercommissioning.

Theageingmanagementprocessesrequirethatthelicenseeregularlyevaluatestheade‐quacyofageingmonitoringplansandthecoverageandeffectivenessofthewholeageingmanagementprogramme.

STUKreviewstheannualageingmanagement follow‐upreportsof the licensees.STUKinspectorsmakeobservationsaboutageingaspartofinspectionsduringoperationandoutagesof theNPPunits, inconnectionwithtests,maintenanceorrepairingworks.Sofar the findings of STUK have been consistent with the ageing management pro‐grammes.

STUK considers that the ageingmanagement programmes concerning theRPVs of theFinnishin‐scopeNPPunitsLoviisa1,Loviisa2,Olkiluoto1,Olkiluoto2andOlkiluoto3areadequate.

AgeingmanagementissuesatNPPunitsLoviisa1and2

There is one ageing management issue that has over the years required significantamountofworkandattentionfromthelicenseesandSTUKaswell.Thisissueistheir‐radiationembrittlementofLoviisaRPVsandthethermalannealingofthecoreareaweldofLoviisa1RPVin1996.TheembrittlementrateofthecriticalcoreareaweldsofbothRPVshastobecarefullymonitoredbythesurveillanceprogrammesaslongastheRPVsareinoperation.Ifthelicenseeplanstocontinueoperatingtheplantunitsafter50years,somemeasuresmaybenecessarytoconfirmsafeoperationoftheRPVs.

Sofaroneindicationhasbeendetectedinalowpressuresafetyinjection(TH)nozzleofLoviisa1RPV.Itmaybecomeanageingmanagementissueifnewindicationswillbede‐tected inothernozzlesof samekind in future inspections.However, it is alsopossiblethattheexistingindicationprovesouttobeamanufacturingdefect.

AgeingmanagementissuesatNPPunitsOlkiluoto1and2

Thedetectedindicationsinthenozzle/safe‐endweldsofsystems312(feedwater)and323(reactorcorespray)maybecomeasignificantageingmanagementissue.Thelicen‐see has decided to continue its work and inspect and, if necessary, repair the corre‐



spondingweldsof systems312,323andalsosystem321(shutdowncooling)atbothunitsOL1andOL2.Sofarone312nozzleandone323nozzleatOL2havebeenrepairedbyamethodthathasnotbeenpreviouslyusedinFinland.

TheoperatinglicenserenewalprocessforNPPunitsOlkiluoto1and2iscurrentlyunderway.ThelicenseeTVOhasappliedpermissiontocontinueoperatingNPPunitsOL1andOL2until60years.InthiscontextTVOhasupdatedthestrengthandfatigueanalysesoftheRPVs. These analyseshavebeenquite comprehensive including almost everythingnecessaryexcept forbrittleand fast fractureanalyses thatarerequired inSection6of[STUK–GuideYVLE.4].TVOhasbeengiventimetocompletetheanalyses.Allrelevantageingmechanismshavetobeconsideredintheanalyses.

Theanalysesperformedso farshowthat themostcriticalsubcomponentsof theRPVsarethefeedwater(312)nozzles.Sothestrengthanalysesanddetectedindicationssup‐port each other. Based on these analyses, the damage risk of all other RPV’s in‐scopesubcomponentsismuchlower.TheanalysesalsoshowthatIGSCCandIASCCcanaccel‐erateageingoftheRPVs.

AgeingmanagementissuesatNPPunitOlkiluoto3

Theunitisundercommissioning.Thelongconstructionperiodhascausedsomeminorageingeffects insomeplantcomponentsbuttherehasnotbeenanyrealageingissuesrelatingtotheRPV.

6 Calandria/pressuretubes(CANDU)

N/A

7 Concretecontainmentstructures

7.1 Descriptionofageingmanagementprogrammesforconcretestructures

7.1.1 Scopeofageingmanagementforconcretestructures

TheconcretestructureswithinthescopeofthisSectionare:

concretecontainmentstructures,withorwithoutaliner,designedtowithstandthe pressure associated with a significant leakage of coolant from the reactorcoolingsystem;and

theconcretestructurethatsurrounds:

o aconcretecontainmentstructureasdescribedinthefirstbullet;or

o a (self‐standing) steel containment designed towithstand the pressureassociatedwithasignificantleakagefromthereactorcoolingsystem.

Thisstructureisoftentheouterwallofthereactorbuilding.

Concretestructures(containmentsandreactorbuildings)aregroupedin threecatego‐ries:



concrete structures in the steel containments and reactor buildings that sur‐roundsteelcontainmentsofLoviisa1andLoviisa2VVER‐typeofPWRnuclearpowerplantunits,

prestressedconcretecontainmentandsurroundingreactorbuildingofOlkiluoto1(built1976)andOlkiluoto2BWRnuclearpowerplantunits,

prestressedconcretecontainmentandsurroundingconcretereactorbuildingofOlkiluoto 3 EPR‐type of PWRnuclear power plant unit (under commissioning,mostoftheloadbearingconcretestructureshavebeencastduring2005‐2011).

Concrete structures in the containment and reactor buildings of NPP unitsLoviisa1and2

TheLoviisaNuclearPowerPlant(NPP)islocatedclosetotheFinnishtownofLoviisa.IthousestwoSoviet‐designed(Atomenergoexport)VVER‐440/213PWRreactors(Loviisa1andLoviisa2),eachwithacapacityof502MWe.

The reactors at Loviisa NPPwent into commercial operation in 1977 (Loviisa 1) and1981(Loviisa2)respectively.TocomplywithFinnishnuclearregulation,Westinghouseand Siemens supplied equipment and engineering expertise. The plant is operated byFortumOyj.

Loviisa1andLoviisa2nuclearpowerunitshaveaself‐standing,weldedsteelcontain‐mentswithahemisphericaldome,cylindricalmiddlepartandconcretefloorslabstruc‐ture(level+10,45),whereinthecentreislocatedthereactorpitconcretestructure.Theconcretefloorslabofthecontainmentandthereactorpitarecoveredwithsteellinerinordertomaintaintheleak‐tightnessofthesteelcontainment.

Theself‐standingsteelcontainmentisanchoredtothereinforcedconcretering‐shapedslabatthelevel+9.60.Thecylinderpartofthesteelcontainmentisweldedasanangleconnectiontoflatsteelanchorflangering,thatisfixedwithpre‐tensionedboltsthrutheconcretefloorslabstructure.

Theself‐standingsteelcontainmentissurroundedbyoutercylindrical0,6mthickcon‐cretewallofthereactorbuilding(innerradius23600mm)andsteelroofstructure.Thebearing structure of the roof is made of steel radial steel girders and tangential ringbeamstructures.

Thefloorslabatthelevel+9.60misaring‐shapedreinforcedconcreteslab,withinnerradius5360mminLoviisa1andradius5410mminLoviisa2andouterradius23750mm.Theslabis1170mmthickinLoviisa1and1370mminLoviisa2.IntheconcreteoftheslabthereareembeddedI‐profilesandlinerplatesoftheslabareweldedtotheup‐perflangeoftheI‐beams.Againstliftcausedbypressuretheboundaryofthering‐shapeslabissupportedtothereactorbuildingwallwithastructure,thatisacombinationofrubberbearing,steelprofileandanchoragerods.

Theinnerconcretestructuresofthesteelcontainmentarereactorpit,floorslab+1,50,boxesandcolumnsonthelevel+10,50m,cylindricalreactorpitwallsandreactorpools,slabson the level+19,30mand+22,20mand+25,40m. Slabon the level+9.60m is



supportedbycylindricalwalls,whichhaveinnerradius5301,12600,20400and23400mm.Additionallythe+19,30slabissupportedbycolumnsontheradius9550mmand16300mm.Thereactorpit(level+0…+10,85)structureisamassivereinforcedconcretecylinder,whichsupportsthereactorandisthefirstbiologicalshield.

Concrete structures are designed according to theory of elasticity (allowable stresses)usingFinnishConcreteCode1971,RIL53C.Limit statedesignhasbeenused in someanalysesusingFinnishConcreteCodeRIL53d.Loadcombinationsinthedesignwereac‐cordingtoACI349‐76.

MaterialsinNPPunitsLoviisa1and2

ThesteelcontainmentisfabricatedusingRautaruukkicarbonsteelRaex385(Loviisa1unit)orRaex305AV(Loviisa2unit).Designofthesteelcontainmentshasbeendoneac‐cordingtoASMECodes;ASMEBPVCodeSectionIIandSectionIIIandSectionVIIIver‐sion1968forLoviisa1containmentandversion1971forLoviisa2containment.

Theconcreteclassof innerstructuresofthecontainmentaregenerallyeitherconcretestrengthclassAK300orAK350(C25/30orC28/35accordingtoSFS‐EN1992‐1‐1).Thecylindricalwallof thereactorbuilding ismadeofclassAK300(C25/30)concrete.TheusedcementinconcretestructureshasbeenmainlyordinaryPortlandcement(OPC)ininnercontainmentandreactorbuilding.Slagcement(OPC+ggbs)hasbeenusedinsomestructuralparts(slabonthelevel+9,60andreactorpit).Theconcretestrengthclassofreactor pit is partly AK90250 (slag cement, strength tested at 90 days age, C20/25,EN1992‐1‐1) and partly AK28300 (portland cement, strength tested at 28 days age,C25/30).

Heavyweightconcrete(density>3500kg/m3)withmagnetiteasanaggregatehasbeenused in structures designed against gamma radiation. Cylindrical biological shield ispartlycoveredbysteelstructurefilledwithheavyweightconcretewithserpentiniteag‐gregate.

RebarsoftheconcretestructureswereribbedbarseithergradeA40HorA40HS(welda‐ble),fy=400MPaandroundbarsA22orA22S(S235).Lapsplicesweremainlyusedbutfor25and32mmbarsalsocouplerspliceswereusedafter1974.WeldedspliceswereusedinthereactorpitoftheLoviisa1unit.

Structures which needed absolute water tightness were covered with stainless steelplatesgradeAISI304L(reactorpit)andgradeOX18H10T(poolsandpits).Thicknessofthelinerswasmainly3mmexceptinthebottomofthereactorpitwherethicknesswas10mm.Theselinerswereeitherprefabricatedliner‐elementsinstalledastheformworkorweldedtothesteelmembersembeddedinconcrete.

Bearingshavebeenused in theexpansion jointsbetweenslabson the level+9,60and+10,50 and reactor pit. These bearings consist of two polyester plates and siliconegrease installedbetweenthem.Thicknessof thebearing is4mm.Betweenthebottomconcretesurfaceandthebearinganeoprenerubbersheetisinstalled.



Concrete structures in the containment and reactor buildings of NPP unitsOlkiluoto1and2

TheOlkiluotoNuclearPowerPlant isonOlkiluoto Island,which ison theshoreof theGulfofBothnia in themunicipalityofEurajoki inwesternFinland.Theplant isownedandoperatedbyTeollisuudenVoima(TVO),asubsidiaryofPohjolanVoimaOy.

TheOlkiluotoplantconsistsoftwoBoilingWaterReactors(BWRs)Olkiluoto1andOlki‐luoto2with880MWeeach.

Unit1achieveditsinitialcriticalityinJuly1978anditstartedcommercialoperationsinOctober1979.Unit2achievedits initialcriticality inOctober1979anditstartedcom‐mercialoperationsinJuly1982.

Olkiluoto1jaOlkiluoto2(OL1andOL2)NPPunitshaveprestressedconcreteInnerCon‐tainments. Containments have been designed against load combinations given in thedocumentVBBDesignCriteria,Rev.H,1977‐11‐15. Structureshavebeendesignedac‐cordingtoFinnishBuildingCodeusing“allowablestressestheory.”

The main parts of the outer shell of the containment are the bottom part (slabs andwalls),cylindricalouterwallandroofstructure.Allthesepartsareconcretestructures.Theouterwallcontainsasteellinerembeddedintheconcrete.Thebottompartconsistsofbasematontherockfoundation,onthelevel‐2.000andacylindershell withinsideradius4500mmonthelevel‐2.000…+1.000withoutsideradialwalls8pcs,whichsup‐portcircularplateonthelevel+2.500.

CylindricalouterwallconsistsofprestressedcylindershellRinside=11000mmonthelev‐el +2.500…+35.000 and circular plate on the level +37.500 and above it walls of fueltanksofwhich thewalls in east‐westdirection areprestressed. Pool structures at thelevel+37.500belongtocontainmentroofslabandpartofthecontainmentwallsverticalprestressingcablesareanchoredtoupperpartofthepoolwalls.

Innerstructuresofthecontainmentconsistthelowercylindricalwallfromlevel2,500tolevel+20,00(RPVpedestal)and theuppercylindricalwall (biologicalshield) fromthelevel+20,00to0,200mfromthebottomofthecontainmentroofstructure.Thereisalsoaringshapeconcreteslabstructurebetweendry‐wellandwet‐well.

The containment wall includes various penetrations such as equipment hatch D2500mmatlevel37,500,personnelairlocksD2500mmatlevel+25.00and‐2.000,pip‐ingD60…D810mmandelectrical cablepenetrationsD324mm, enabling connectionswiththeotherbuildingsanddifferentsystems.

MaterialsofNPPunitsOlkiluoto1and2

The strength class of the concrete used for the containment structures was AK40(C32/40 according to EN1992‐1‐1). In allmassive structures low heat cement (KolarilowheatcementinOL1andLohjaggbs/Portland‐cementinOL2)hasbeenused.Forthereinforcement Swedish ribbed bars grade Ks40 or Ks40HS (weldable, fy= 400MPa)wereused.A lapsplicewas thepredominantmethodused forsplicingrebarsbutalsosomecouplers(mechanicalsplices)wereused.



TheprestressingsystemwasVSL(VorspannsystemLosinger) inOL1unitandBBRVinOL2unit.Tendonductsweregroutedwithcementbasedgrout. Tendonsused inOL1unitandOL2unitaregiveninthefollowingTable7.1.

Table7.1.TendontypesusedinOlkiluoto1andOlkiluoto2nuclearplants.

Olkiluoto1 Olkiluoto2prestressingsystem VSL(VorspannsystemLosinger) BBRVtendonsofcontainment 122horizontal19BridonSupa

LRwiresd=13mm40verticalcableseachwith19wiresand18verticalcableswith31wires

122horizontalcableseachwith72D6mmstrands35verticalcableseachwith72D6strandsand12verticalcableseachwith115D6strandsand14verticalcableseachwith78D6strands

tendonsinpoolwalls 48horizontal cableseach with19wires

48verticalcableseachwith72D6strands

anchors VSLtypesE5‐19(for19D13)andE5‐31(for31D13)

activeBBRV typeR‐268(for78D6)verticalandhorizontalcablespassiveBBRVtypeR115(for115D6)

specimenbeams BBRVtypeR‐238(72D6)

Leak‐tightnessofthecontainmenthasbeeninsuredwith5mmthickcarbonsteelplateHII/WN:o1.0425(DIN17155)linerthathasbeenembeddedabout200‐250mmintheconcretestructure.Thicknessofthelineris8mminthewallstructureofcondensationpools(wet‐well).Theinnerconcrete layerprotects lineragainstmissiles , temperaturevariationandcorrosion.Thelinerisnotanchoredtotheconcreteinthecircularwalls.

Fuel pools and condensation pools are covered with 3mm thick stainless steel platestructure.

Concrete structures in the containment and reactor buildings of NPP unitOlkiluoto3

ThebasematoftheunitOlkiluoto3wascastedJanuary‐October2005.Olkiluoto3unit,anEPRreactor,isstillundercommissioning.

ThecontainmentoftheOlkiluoto3EPRprojectisadouble‐wallstructurefoundedonabasemat.Theinnercontainmentshell isaprestressedcontainmentwithsteel linerim‐plementedon the innersurface including thebasemat, thus formingacontinuoussur‐face.

Intheprimarycontainment(insideinnershell)thewholeReactorCoolantSystem(RCS),theIn‐ContainmentRefuellingWaterStorageTank(IRWST),thesteamgenerators,andpartofthemainsteamandmainfeedwaterlinesaresituated.Theinsideradiusofcylin‐derpartinprimarycontainmentis23,40m.Designpressureoftheprimarycontainmentis5,3barabsolute.

Theprimarycontainmentiscomposedof



a6mmsteelliner,

the liner anchorage system to the concrete,welded to the liner (stiffeners andstuds),

aprestressed1,30mthickcylindricalwall,with1or2layersofhorizontaland1layerofverticalprestressingtendons,

aprestresseddomeapproximately1.00mthick,with2setsofprestressingten‐dons

Thecontainmentlinerconsistsofsteelplatesontheinnersurfaceandananchoragesys‐temontheconcretesidewithwhichthecontainmentlinerisconnectedtotheinnercon‐tainmentwall.

Sleevesarewelded into thecontainment linerpenetrating the innercontainmentwall,whichareusedforpipingsystems,HVACandelectricalpenetrations.Steelworksinsidethe containment arewelded onto anchor plates,which are also integrated part of thecontainmentliner.TheanchoringsystemofthecontainmentlinerconsistsofL‐profiles(L‐anchors),whichareweldedontothesteelplatebycontinuousfilletwelds,limitingacontainment liner field. Within the containment liner field headed studs are welded,whicharearrangedinasquarepatternof150mmor75mm(dome).

Thecontainment liner includesvariouspenetrationssuchasequipmenthatch,person‐nelandemergencyairlocks,pipingandelectricalcablepenetrations,andthefueltrans‐fertube,enablingconnectionswiththeotherbuildingsanddifferentsystems.

Eachhorizontaltendon(totally119)wasjackedoverfullroundof360˚frombothendsononeverticalrib(threeribsintotal).Pureverticaltendons(totally47)werejackedbytheloweranchoragelocatedinprestressinggallerybelowthebasemat.Dometendons(totally104)werejackedonbothendslocatedinprestressinggalleryanddomebelt.

Theoutercontainmentshellisareinforcedconcretestructure.ItguaranteesprotectionagainstexternalhazardssuchasairplanecrashAPCandexplosionpressurewaveEPWandwithstandstheloadsofasafe‐shutdownearthquakeSSE.

Thetwoshellsareseparatedbytheannulus,1.80mwide,placedundersubatmosphericpressureinordertocollecttheleaksthroughtheinnercontainmentandfilterthembe‐forereleasetotheenvironment.

The basemat is 3,15m thick reinforced structure. The steel liner is implemented be‐tweenthecontainmentbasematandthebasematofthereactorbuildinginternalstruc‐ture,whichensuresanyreleaseofradioactivitytothegroundwater.Acircumferentialpre‐stressingtendongalleryofverticaltendonsoftheinnercontainmentissituatedun‐derneaththebasemat.

Theinternalcivilstructuresinsidetheinnercontainment,whichhousethereactorcool‐antsystemrestonthecontainmentbasemat.Thisinterfacebetweeninternalstructuresandcontainmentbasematisprovidedwithaleaktightsteelliner.Thesteellinerisim‐plementedontheinnersurfaceoftheinnercontainmentshell.



MaterialsinNPPunitOlkiluoto3

TheconcreteusedfortheinnercontainmentprestressedshellisstrengthclassC50/60(EN1992‐1‐1)andallotherconcretestructuresarestrengthclassC32/40.InallmassivestructuresbinderwithPortlandcementand50‐70%slag(ggbs)hasbeenused.

For reinforcementhot rolled ribbedbars gradeA500HW(Finnish standardSFS1215)and A500SAS (Type Approval YM156/6221/2006, Stahlwerk Annahütte) were used(strengthfy=500MPa,ductilityclassB).CouplersofthereinforcementwereeitherLen‐ton(EricoB.V.)orSAS500(StahlwerkAnnahütte).Couplershavebeenusedbothintheprimary containment and in the secondary containment i.e. outer wall of the reactorbuilding.

TheprestressingsysteminOL3containmentisFreyssinetwithEuropeanTechnicalAp‐provalETA‐06/0226.Eachtendonhas54T15,7lowrelaxation(class2)strands.Forpre‐stressingsteelmaterialwithstrengthfp0.1k/fpk=1653MPa/1860MPahasbeencho‐sen. The strandswere 7‐wire low relaxation type Y1860S7‐15,7 fabricated byArcelorMittal.Theappliedcodeforpre‐stressingsteelmaterialwasprEN10138‐3.

The anchors of the tendons are Freyssinet “C” Range Prestressing Anchorages with55C15 trumplates. Trumplates aremade of spheroidal ductile cast‐iron grade EN GJS500‐7accordingtoFrenchstandardNFEN1563.

Mostofthetendonductsaregroutedwithcementbasedgrout,exceptthefourpurever‐ticaltendonsfittedwithdynamometerswhoseductsareinjectedwithwax.ThecementgroutcomplieswithENcodesEN445,EN446andEN447.TheinjectionwaxisInjectELFCP‐HFfulfillingETAG013requirements.

TheinterioroftheInnerContainmentwasequippedwitha6mmthickcontainmentlin‐er.Thematerialof the linerwas finegrainedsteelP275NL2(EN10028) fabricatedbyRautaruukkiOy.Thelinerconsistsofthesteelplates,thesystemofcircumferentialandmeridianstiffenersandthelinerstuds.Thestiffenerprofilesandthestudswereembed‐dedintheconcrete.TheanchoringsystemofthecontainmentlinerconsistsofL‐profiles,whichwerewelded onto the steel plate by continuous filletwelds, limiting a contain‐mentlinerfield.Withinthecontainmentlinerfieldheadedstudswerewelded.Thema‐terialofL‐anchorswasS235JRG2andheadedstudsweremadeofS235J2G3.

TheanchorplatesweremadeofcarbonsteelS235incompliancewithDINEN10025.

The sleeveswelded into the containment liner andpenetrating the inner containmentwallweremadeoffinegrainedsteelP275NL2andP355NL2(EN10028).

TheReactorCavity,CoreInternalStorageandTransferCompartmentandInstrumenta‐tionLancesStoragePoolsintheReactorBuildingwereequippedwithpoolliner,follow‐ingstainlesssteeltypeswereusedinthefabrication

martensiticstainlesssteelbarsX20Cr13(1.4021)QT700

martensiticstainlesssteelX17CrNi16‐2(1.4057)QT800



austeniticstainlesssteelX6CrNiTi18‐10(1.4541)

austeniticstainlesssteelX6CrNiNb18‐10(1.4550)

austeniticstainlesssteelX6CrNiMoTi17‐12‐2(1.4571)

austenitic stainless steels with steel grades A2, A3, A4 and A5 and strengthgrades50,70and80forconnectingelements

stainlesssteel1.4301forheadedstandardstuds

forheavy loaded anchorsnon‐alloy steel (1.0038, 1.0570)wasused insteadofthestainlesssteel

The concrete surfaces of the IRWST‐ tanks (In‐Containment RefuellingWater StorageTank)arecoveredwith5mmthickausteniticsteelliner(1.4571).Linerplatesareweld‐ed tostainlesssteelsupports (U‐profiles1.4571andheadedstuds1.4303) ,whicharefixedtotheconcretewithferriticsteelanchors.

7.1.2 Ageingassessmentofconcretestructures

ThisNARdescribes theageingassessments foreachNARexample i.e.Loviisa1and2,Olkiluoto1and2andOlkiluoto3.

ThisNARdescribestheoutputsfromtheageingassessmentincludingthefollowing:

a)Ageingmechanismsrequiringmanagementandidentificationoftheirsignificance;

Indescribingtheseoutputs,theNARexplainshowthefollowingareused:

•KeystandardsandguidanceusedtopreparetheSSCspecificAMP, includinga listofthemaindocuments;

•Keydesign,manufacturingandoperationsdocumentsusedtopreparetheSSCspecificAMP;

•R&Dprogrammes, bydescribing theobjectives, contents and results of programmesmanagedby:

oLicensees;

oIndustryorotherrelevantbodies;

•Internalandexternaloperatingexperience,bydescribingwhyandhowtheseexperi‐enceshavebeentaken(ornot)intoaccount.

ThesafetyrelatedconcretestructuresinFinnishnuclearpowerplantandthedegrada‐tion phenomena in those structures were identified in the report VTT‐R‐02323‐08[Vesikari2008]reportsundertheauspicesofSAFIR2010researchprogramme.



NPPunitsLoviisa1and2AgeingmechanismsFollowing concrete structures in Loviisa 1 and2 containments and reactor buildings ,theirageingmechanismsandidentificationoftheirsignificancetotheplantsafetywerelistedinSAFIR‐researchprogrammeinthereportVTT‐R‐02323‐08[Vesikari2008]:

Agingmechanismsofconcretestructuressupportingthesteelcontainment:

T‐weldoftheanchorringatlevel+9,60,possiblelamellartearing Ringlayer+9,60,corrosionofthepre‐tensionedfasteningboltsofthesteelcon‐

tainmentthrutheconcretefloorslabstructure,service‐lifemaintainedwithin‐spectionprogramme,considerableeffectstotheplantsafetyifperiodicinspec‐tionsandchangeofboltsarenotconducted.

Ringlayer+9,60,agingoftheslipbearing(glassfibrereinforcedpolyesterstruc‐

ture),damagescausedbythermalmovementofthestructuresaboveandbelowthebearing,considerableeffectsonsafety.

Columns,walls,outercylinderofthereactorpit,waterleakage,crackingcausing

corrosion,moderateeffectonsafety.Agingmechanismsoftheinnerconcretestructuresofthesteelcontainment:

Innercylinderofthereactorpit,reactorpitandsupportingareaofthereactor,degradationofconcreteduetoelevatedtemperatureandirradiation,considera‐bleeffectsonsafety.

Poolstructures,leakageofthepoollinerduetocorrosion,moderateeffectonsafety.

Level+10,50surfaceslab,damageinthefloorsurfacingandimmersionofthe

leakagewaterintoconcretecausingcorrosion,minoreffectonsafety.

Areaoftheprincipalcircuitvalve,locallyhightemperature,crackingofconcrete,minoreffectsonsafety.

Concretestructuresoftheicecondenserandcranewall,moistureandtempera‐

turevariations,crackingofconcrete,minoreffectsonsafety.

Columnsandwalloftheevaporatorroomandlevel+25,40slab,,temperaturevariations,crackingofconcrete,minoreffectsonsafety

CylinderwallR=9550(radiationshield)madepartlyofmagnetiteconcrete,irra‐

diationandhightemperature,degradationofconcrete.



Level+19,30(leveloftheevaporatorsupport),variationintemperature,crack‐ingofconcrete,minoreffectsonsafety

Agingmechanismsofthereactorbuildingoutershell:

Concretecylinderwallcladdedwithcorrugatedsteelplate,moisturestress,deg‐

radationandcrackingofconcrete,frostdamage,corrosionofreinforcement,mi‐noreffectsonsafety.

Rockanchorsofthecylinderwall,reductioninalkalinityaroundrockanchorsandcorrosion,reductioninanchoragecapacity,minoreffectonsafety.

Roofsupportedbysteelarchesandframes,corrosion,lossoftherooftightness,minoreffectsonsafety,diminishesthepressurebetweentheoutershellandsteelcontainment.

Partofthecylinderwallwithoutcladding,weathering,frostattackonconcrete,corrosionofreinforcement

NPPunitsOlkiluoto1and2AgeingmechanismsFollowingconcretestructuresinOlkiluoto1/2containmentsandreactorbuildings,theirageingmechanismsandidentificationoftheirsignificancewerelistedinSAFIR‐researchprogrammeinthereport[VTT‐R‐02323‐08,Vesikari2008].Theagingmecha‐nismspresentedbelowaregatheredeitherfromtheVTTreportorfromtheinformationgivenbythelicensee:AgeingmechanismscertainstructuresoftheContainmentBasematonrock,topsurfaceoutsidesteelliner,tendongallery

Bottomsurface,chlorideinitiatedcorrosionofreinforcement,crackingandspallingofconcrete,onlyminoreffectsonsafety

Topsurfaceoutsidesteelliner,tendongallery;corrosionofprestressingtendonanchors,reductioninprestressingforce,reductioninpressurebearingcapacityofcontainment

Topsurfaceinsidethesteelliner,corrosionofsteelliner,reductioninleaktightness,radioactiveleakageduringaccident

In‐situcastbottompartsinsideandoutsidesteelliner

Corrosionofsteelcausinglinerreductioninleaktightness,radioactiveleakageduringaccident

Leakagethroughthestructuretothedrywell,degradationofconcrete,riskontheloadbearingcapacityoftheinnercylinderwall



Highrelativehumidityofconcrete(OL1),corrosionofsteellinerandreductioninleaktightness

Pre‐stressed,slip‐formcastoutercylinderwall

Corrosionofprestressingtendons,reductioninprestressingforce(losses),re‐ductioninpressurebearingcapacityofcontainmentduetolargerdeformationsthanexpected

Relaxationofprestressingtendonsandshrinkageandcreepofconcrete,reduc‐tioninprestressingforce,reductioninpressurebearingcapacityofcontain‐mentduetolargerdeformationsthanexpected

Corrosionofsteellinerreductioninleaktightness,radioactiveleakageduringaccident

Repairoffiredamagesbyshotcreting(OL1),detachmentofshotcretelayer,riskonloadbearingcapacityandleaktightnessoftheoutercylinderwall

Slip‐formcastinnercylinderwall

Leakagethroughthestructuretothelowerdrywell,degradationandcrackingofconcrete,riskontheloadingcapacityoftheinnercylinderwall

Biologicalshield

Innersurface,crackingofconcrete,reductioninload‐bearingcapacity,reductioninthestabilityofthereactorsupport

Outersurface,irradiationandhightemperature,degradationofconcrete,reduc‐tioninthestabilityofthereactorsupport

Roofstructureandnon‐prestressedpoolstructures

Corrosionofpoolliners,reductioninleaktightness,radioactiveleakageduringaccident

Leakageofpoolliners,degradationofconcrete,onlyminoreffectsonsafety.Wet‐wellpoolleakagecanbeaproblemforcontainmentlinercorrosion.Poolleakagecanalsostartconcretedegradation.

Intermediatefloor

Upperandlowersurface,crackingofconcrete,,reductioninleaktightness,re‐ductioninleaktightness,riskonpressuresuppressionduringaccident.

Leaktightnessreduction:moving jointpolymericseal, transport jointpolymericseal,electricalpenetrationsandvicinityconcretecracking,doorpolymericseals

Prestressedpoolstructures Relaxationofprestressingtendons,reductioninprestressingforce,riskondis‐

placementsandcrackingandleakageinpoolstructure Creepofconcrete,reductioninprestressingforce,riskondisplacementsand

crackingandleakageinpoolstructure



Loadchangesliketemperaturechanges Gateoperations

Containmentandreactorbuildingdifferentverticalmovement

Verticalmovementjointsweretootight,theyhavebeenenlargedAnchoringareasforprestressingcables

Concretecovercrackingduetoconcreteshrinkage,riskforcorrosionAirlock

SealmaterialageingLinercorrosion

Constructiontimeimperfections,concretecleanness,possibleproblemsduetolinersurfaceorganicimpurity(gloves,thermalinsulationetc.)

NPPunitOlkiluoto3

AgeingmechanismsFollowingdescriptionoftherelevantagingmechanismsinContainmentandReactorBuildingofOlkiluoto3nuclearpowerplantisbasedontheAREVAreportPESS‐G/2011/en/1104.LeadingAgeingMechanism,CreepandShrinkageoftheConcreteandRelaxationofthePre‐stressingSteelWithrespecttotheloadbearingfunctionoftheInnerContainmentconcretewallandtheleaktightnessfunctionofthelinertheleadingageingmechanismsarecreepandshrink‐ageoftheconcreteinconnectionwiththerelaxationofthepre‐stressingsteel.Thecreepandshrinkageoftheconcretetogetherwiththerelaxationofthepre‐stressingsteelleadtoadecreaseofthepre‐stressingforceandadecreaseoftheconcretecom‐pressionstress.Ifthistimedependentprocesswouldlastunlimiteditwouldfinallycausetensilestressesandcracksintheconcreteundercertaindesignloadconditions.ChemicalAttackandCarbonationoftheConcreteAstheOuterContainmentofOlkiluoto3ReactorBuildingprovidesacomprehensivepro‐tectionagainstdetrimentalweatherconditionsduringtheNPPlifetime,chemicalattackandcarbonationoftheconcretethereforecouldtakeplaceonlyforshorttime.Thesemechanismscan’tcausearelevantageingoftheconcretestructure.ChemicalAttackandCorrosionoftheReinforcementBarsandthePrestressingSteelChemicalattackandcorrosionofthereinforcementbarsandpre‐stressingsteelhavebeenlimitedbysuitableprovisionsinfabricationofconcreteandinjectiongrout:



thechlorideandsulphatecontentofthewaterandofadmixturesinconcreteandinjectiongroutislimited

theaccessofwaterandairtoreinforcementbarsandpre‐stressingsteelisaverted

afterconcretingduringthecontainmentlifetime.Possiblechemicalreactionsarestoppedshortlyafterconcreting,becausethereservoirofwaterandairislimitedattheconsideredlocations.

airbubblesorvoidsbetweenthegroutedpre‐stressingstrandsareimpededbyproperlyinjectionprocedureswhichhavebeenapprovedbyfull‐sizemock‐upsoncurvedtendonductsbeforetheapplicationontheInnerContainmentwall.

chemicalattack(sulphate)onthebasematistakenintoaccountbyusingsul‐phateresistant(70%ggbs)‐slagconcreteinthebasematandbyprotectingtheundergroundboundariesofthebasematwithbitumencoating.

CorrosionofthePre‐stressingAnchorageSteelParts

Corrosionofthepre‐stressinganchoragesteelpartsispreventedbytheinstalla‐tionofgroutingcapsandinjectingofgroutinsidethecapsafterpre‐stressingthetendons.

AgeingMechanismsfortheLiner

Becausethecontainmentlinerisconnectedrigidwiththeconcretewall,therel‐evantageingmechanismfortheconcretewallislinkedtothecontainmentlinerregardingtheleaktightnessfunction.Thelinerisperformedwithagridofstudsandstiffenerswhichareembeddedintheconcreteofthecontainmentwall.Thedeformationoftheconcretewallduetocreepandshrinkageisdirectlytrans‐ferredtothelinerplatebythelinerstudsandthestiffeners.

AtthebeginningoftheNPPlifetimethelinerisalreadyundercompressiondue

totheprestressingoftheInnerContainmentwall.Thestrainsderivingfromcreepandshrinkageoftheconcreteincreasethenegativestrainsoftheliner.Theresultingnegativelinerstrainscancauseablisteringoftheliner.Althoughtheleaktightnessofthelinerisensuredevenwithblistersthisdeformationisre‐gardedasanageingmechanism.

Linercorrosionisavoidedbythecoatingbothsurfacesoftheliner.Theliner

coatingisnotpartoftheAMP,becausetheremovalandreplacementofthecoat‐ingisanywayinscopeofthemaintenanceprogramme.

Belowthelevel4.30mthelinerisembeddedintheconcreteandthereiscoating

atbothsurfacesoftheliner,replacementofcoatingormaintenanceinspectionsarenotpossible.

7.1.3 Monitoring,testing,samplingandinspectionactivitiesforconcretestructures

TheNARshoulddescribethemonitoring,testing,samplingandinspectionactivitiesforeachspecifiedelementforeachNARexample.



Themonitoring, testing, sampling and inspection activities performed by the licenseeandactivitiesperformedby‘thirdpartycertificationorganizations’shouldbedescribedincluding:

a) Descriptionofactivities;

b) Frequencies;

c) Acceptancecriteria.

Indescribingtheseoutputs,theNARshouldexplainhowthefollowingareused:

Programmes formonitoring and trending including testmethods available foruseinperforminginspections;

Keyfeaturesoftheinspectionprogrammes;

Surveillanceprogrammeswhereappropriate;

Inspectionhistoryidentifyingtrendsandprogressivedeterioration;

Anyprovisionsforidentifyinganyunexpecteddegradation.

NPPunitsLoviisa1and2

Description of monitoring and inspection activities in Loviisa 1 (LO1) and Loviisa 2(LO2)nuclearpowerplantsisbasedonthefollowingguidesofthelicensee(Fortum);Y‐05‐00005“Periodicalcivilengineeringinspections”.

Pressureandleak‐tightnesstests

Pressuretesttothewholesteelcontainmenthasbeenconductedtwiceduringcommis‐sioning phase.The testpressureswere85,8kPa inLO1and for78,9kPa inLO2con‐tainment.InLO1andLO2NPPleaktightnessofthesteelcontainmenthasbeenverifiedwithpressure testswithdesignpressure(1,7barabs)every4thyear,afteryear2015thetestintervalhasbeenincreasedto8years. .AcceptancecriteriaoftheleakagerateforLO1andLO2is0.2Vol‐%per24hours.

Periodicalinspections

Guide Y‐05‐00005 defines those reinforced concrete structures and steel structures,whichareunderthescopeofperiodicalcivilengineeringinspectionsanddescribestheinspectionactivitiesforeachstructure.

InspectionsareconductedasvisualinspectionsusingalsoNDE‐methods.Inspectionsareconducted as a general inspection for each building according to the inspection cardsandobjectivesandcriteriagivenintheAppendicesoftheGuideY‐05‐00005.Forsafetycriticalstructures inspectionareconductedonceayear,otherstructureseverysecondyear.



Cracksthatarecriticalforsafety,agingorstructuralintegrityarerecordedintheinspec‐tion cards. Damageswhich need specialmonitoring are photographed,measured andmarkedonthestructure.

ClassificationofinspectionfindingsinNPPunitsLoviisa1and2

InspectionfindingsofareclassifiedbasedonthecriticalityoftheSSCtothenuclearandradiation safety and theurgencyof remedial actions.Theprimarydigit of thedamageclassification code (for example A1) describes the effect to the nuclear and radiationsafety:

Class0Criticalstructuretothenuclearandradiationsafety(loadbearingstruc‐turesofthereactor,radioactivefuelpoolsetc.).

ClassACriticalstructuretothesafety,limitingplantlife.

ClassBStructurehaseitherahighimportancetoheathandsafetyorisanobjecttochemicalorheatattackradioactivecontaminationetc.

ClassCStructureoritsfunctionalityanddamagedonothaveanimmediateeffecttothesafety.

Thesecondarydigitoftheclassificationdescribestheurgencyoftheremedialactions:

0:Aseriousdamagewhich threatens loadbearingcapacity, stabilityandstruc‐turalintegrityofthestructure

1: A seriousdamagewhichdoesnotprevent theuseofpowerplantunit.Thedamageissituatedsothatitthreatenshealthandsafetyortheprogressofthedamagemaycausealossofloadbearingcapacity,stabilityandsustainabilityofthestructure

2:Asignificantdamagebuttheprogressofithasonlyaminoreffecttotheloadbearingcapacity,stabilityandstructuralintegrityofthestructure

3:Thedamageisminorandtheprogressofthedamageisslow

4:Remedial actions increasehealth andsafety, decontaminabilityorotherwiseimprovestructuralfunctionality.

Theclassificationsandcolourcodeusedintheinspectionreportsarefollowing:

00:Aseriousdamagewhichpreventstheuseofpowerplantunit,hastobere‐pairedbeforetheendoftheoutage(red)

A1:Mustberepairedimmediately,structureiscriticaltosafety(red)

B1:Must be repaired immediately, damage that needs immediate remedial ac‐tions(red)

A2:Mustberepairedinthenearfuture,structureiscriticaltosafety(orange)



B2:Mustberepairedinthenearfuture,damagethatneedsremedialactions(or‐ange)

B3:Mustberepairedinduetime(yellow)

C1:Norepairneeded,progressofthedamageIsmonitored(green).

IntheAppendicesoftheGuideY‐05‐00005inspectiontargetsandcriteriaaregiven.ForexampleintheAppendix4thefollowingcriteriafortheinspectionofthereactorbuild‐ingarepresented:

Structuresbelowlevel9,60:

Reinforced concrete slab at level +9,60, visual inspection of upper andbottomsurfacecracksandotherdamagesarerecorded,concretearoundpenetrationsisinspected, slab’s connection to the outer cylinderwallmust be intact, possiblecracksarerecorded.

Loadbearingoutercylinderof thereactorpit fromlevel+9,60to therocksur‐face,possiblecracksaremeasuredandrecorded.

Columnsandwallsbelowlevel+9,60,possiblecracksaremeasuredandrecord‐ed.

Equipmenthatches1R0301(LO1)and2R0301(LO2),pressurebearingstructure,possiblecracksintheconcretearoundthehatcharemeasuredandrecorded.

Emergency water pools (boron water pools) bottom slab and walls, possiblemarksofleakagearerecorded

Innerstructuresofthereactorbuilding:

Pressureresistantdoorsandhatches,thefunctionalityofthedoorsandhatchesarechecked,andtheconditionofthesealantsarevisuallyinspected,functionali‐tyofthelockingofreinforcedconcretehatcheslockedwithboltsisinspected

Reinforced concrete and steel structures (functionality), general inspection ofcracksonallconcretesurfaces,over0,2mmwidecracksarerecorded,connect‐ing steel beam of the bottom structure of ice condenser and slab on the level+19,30isinspected

Heavycomponents;ringbeamofthepolarcraneisvisuallyinspected,conditionoftheconcretearoundthesteelsupportsofthemaincoolantpumpsisinspected.Concretestructurearoundtheopeningsontheconcreteslabisinspected.

Outercylinderwallandtheroofstructure;theroofstructureanditsconnectionsto the cylinderwall are inspected visually. Damages causing deterioration thesteelbeamsortheirconnectiontothecylinderwallorincreasingtheleaktight‐nessofthestructurearerecorded(rustofthesteelstructures,locationandcon‐ditionofneopreneslabs,conditionofbolts,possiblewaterleakage)



Expansion/movementjoints

Elasticityandintegrity(airpressure,leaktightness)oftheexpansionjointseal‐antsandtheconditionofconcretestructurearoundtheexpansionjointsarein‐spected(onthelevel+10,50,radiusR=4800mmandonthelevel+25,40radiusR=4500mm). Damages of the expansion joint sealants are classified either “2,repairinaduetime”or“3,nourgentneedforrepair”.

NPPunitsOlkiluoto1and2

According to theAMPGuide107835of the licenseedescriptionofmonitoring and in‐spectionactivitiesinOL1andOL2NPPunitsisbasedonthefollowingAMPsandguidesofthelicensee(TVO);

Deformationmeasurements

AMP113016Strainmeasurementsofthecontainment

AMP104730Deformationsofthecontainment,

Temperaturemeasurements

AMP111402Temperaturemeasurementsofconcretestructures

AMP111407Temperaturemeasurementsoftheturbinefoundations

Crackingsurveillance

AMP104733Surveillanceofthecracksofthecontainment

AMP111407Surveillanceofthecracksoftheturbinefoundations

Leakandmoisturedetection

AMP111409Containmentconcretehumidityfollow‐up

AMP111555Leakagedetectionoffuelandwetwellpools

Leak‐tightnesstestsofthecontainment

AMP104533Leak‐tightnesstestsofthecontainment

Surveillanceofthecontainmentmaterials

AMP111556Surveillanceofthepropertiesoftheprestressedtendons

AMP104756Surveillanceoftheconditionofconcreteincontainment

AMP112739Surveillanceoftherubbersealmaterialsofthecontainment

Periodicalinspectionsofthebuildings



Periodicalinspectionsandmeasurementsareconcentratedinbuildingsandstructures,thatareimportanttotheuseandsafetyoftheNPP´s.Thesebuildingsandstructuresareforexample;thecontainment,reactorbuilding,turbinefoundations,seawaterchannels,poolstructuresandsurfacesoftheradioactiverooms.

Followingitemsareinspectedandrecorded:

crackmapping,cracksoftheconcretestructures

moistureandcorrosiondamages

malfunction of different types of steel structures for example doors, hatches,penetrationsandanchorageplates

changesinthequality,sustainabilityandcolourofthepaintings

conditionoffacadesandroofstructures

waterleakagesintheundergroundspaces

Thecorrectfunctionalityofbuildingsandcriticalstructuralmembersisensuredwithin‐spectionsandmeasurementsdoneyearly.Inspectionandmeasurementresultsaregath‐ered,analysedandreported.Whenneededremedialactionsarestarted.

Frequencyoftheinspections,gradedapproach

Frequency of the visual inspections of the buildings is graded according to the im‐portance of the building, for example; containment is inspected every year, reactorbuilding,turbinebuildingandwastebuildingeverysecondyear,maincontrolroomeve‐ryfourthyear.

InthemaintenanceprogrammeAMP103459inspectedroomsareclassifiedaccordinginafollowingway(inspectionfrequencyishighestinthehighestClass);

Class1Sroomsinthecontainment

Class I A spaces with high radioactivity on the ground floor, where is lots ofmaintenancework

ClassIIAotherspacesofhighradioactivity,wherefrequentmaintenanceworkisneeded

ClassIIB,spaceswithratherlowradioactivity(“greenspaces”)oruncontrolledspaces,wherefrequentmaintenanceworkisneeded

Class IVB, other ”ordinary” controlled spaces like switchgear or I&C cabinets,ventilationshaftsandrooms,corridors,staircases

ClassEspaceswithconstantradioactivity,likeradioactiveshafts



Frequencyoftheinspectionsisdependentontherapidityofthedamage,importanceofthespaceorstructure.Theconditionoftherooms(spaces)iscontrolledinone,twoorfouryearscyclesdependingontheimportanceofthespace.

In theAppendicesof theAMP103459 inspection targetsandcriteriaaregiven.Forex‐ample in the Appendix 3 the following criteria for the inspection of the containment(ClassS,frequencyonceayear)arepresented.Followingitemsarechecked;

Loadbearingandotherimportantstructures(cracking,spalling,deformations)

Surfacesofthespaces(cracking,flaking),acceptancecriteria:

o surfacesshallbecleanandshiny

o nocracksareallowed

o adhesionofthepainting(tensiletests)

Movementjoints(cracking,hardening)

o movementjointmaterialsshallbeinperfectcondition

Penetrations(cracking,spalling,leaktightness)

o fireresistantpenetrations

o watertightpenetrations

o pressureresistantpenetrations

Steelplatformsand–stairs(deformations,fixings)

Doorsofthewetwell,lockingandseals

CurrentmonitoringsystemoftheNPPunitsOlkiluoto1and2containments

Inthemiddleofthe‐90’scontainmentstructuralmonitoringwasupdatedinOlkiluoto1(OL1)andOlkiluoto2(OL2)nuclearpowerplants.Concretehumiditysensorsystemanddisplacement system (mechanical clocks) between reactor building and containmentstructurewereinstalled.Lateralso"eddycurrent"basismovabledeformationmonitor‐ing systembetween containment inner andouter structureswere installed,whichareusedduringpressuretests.Concretecrackmappingwasstartedasapartof thevisualconcretesurfaceinspectionaccordingtoASMEXI.

InOL1andOL2displacementsbetweenthecontainmentandreactorbuildingframeanddisplacementbetweenthemiddlefloorslabandcontainmentwallaremonitoredwithmeasurements. Displacements are recorded once a year during outages. During leaktightness testsdisplacementare recordedbefore the test,during the test inmaximumpressureandafterthetest.Displacementmeasurementsarerecordedusingmechanicalclocks(10pcs).Duringleaktightnesstestdisplacementsbetweenthemiddlefloorslab



ofthecontainmentandrooftopofthecontainmentaremeasuredconstantlywith"eddycurrent"basismovabledeformationmonitoringsystem.

Reinforcement strains in the containment are monitored. OL1 and OL2 containmentoriginalstructuralmonitoringincludes26straingaugesandtemperatureprobes,whicharefixedtoreinforcement.Straingaugesarereadonceayear.Duringleaktightnesstestthestraingaugesreadingsarerecordedonpressurelevels100,150,250,350,400,350,250,150,100kPaabs.

Temperatureoftheconcretestructuresofthecontainmentaremonitoredusingembed‐dedthermocouples.Standardthermometrydevices(10pcs)areusedtomeasureambi‐enttemperatureinsideandoutsidethecontainment.Concretetemperaturesaremeas‐uredwith13thermocouples,whichareembeddedintheconcreteinthevicinityofthestrain gauges. Temperatures aremeasured once a year during outages or during leaktightnesstestsatthesametimeasstrainsarerecorded.

Concretecracksare inspectedonceayearduringoutagesand leaktightnesstests.Thepurposeof the inspection is toobservepossiblenewcracks andmonitor thedevelop‐mentoftheexistingcracks(length,width).Duringnormaloutagesallconcretesurfacesareinspectedandtheexistingcrackswhichareundersurveillanceareanalysed.Duringleak tightness test concrete surfaces are inspectedbeforepressurization,duringmaxi‐mumpressureanduponcompletionofdepressurization.

Acceptance criteria of the cracks is based on the Finnish Concrete Code BY50[BY50,2012].ExposureclassofthecontainmentisXC3.RequirementsforthecrackingofconcretearegiveninBY50table2.16a;crackingshallbelimitedto0,2mmunderlong‐termloads.Whencrackswidthsareover0,20mm,repairplanisdone,ifdangerforcor‐rosionofrebarsortendonsisexpected.

Containment prestressing forces have been tested and analysed with 3 m long testbeamseveryfiveyearsduringfirst25years.

Concretequalityhasbeentestedwithoriginaltestcylinderseveryfiveyearsduringfirst25yearsandafterthatbytakingcoredrillsandwithNDTmethods.

Allcontainmentwaterpools includewater leakagemonitoringsystem.Visualconcretesurfaceinspectionhasbeenperformedyearly.

FuturemonitoringsystemofOL1andOL2containment

Itisnotpossibletodirectmeasureprestressingforcesduetocementgroutedtendons.

U:S. Regulatory Guide 1.90 [newDraft Regulatory Guide DG‐1197, April 2011] part Bgivesanalternativewaytodetermineprestressingforcesforgroutedtendons;“Measurecontainmentdeformationsduringpressuretestandcalculateprestressingforcesaccord‐ingtodisplacements.”Thismethodhasbeenused inOL1andOL2first timeandsince2002 (15 years ago). Displacement between containment and reactor building weremeasured with measurement clocks in the movement joints and deformations werestudiedusingafiniteelementmodelofthecontainmentstructure.



TwokindsofFE‐modelswereused;

linearmodel,wherehalfofthecontainmentwasmodelled(180degreemodel)

wedgemodelthatutilizednon‐linearreinforcedconcreteelementswithcapabil‐ity to model concrete cracking and non‐linear reinforcement response. FE‐programmeANSYSwasusedintheanalysiswithsolid65elements.

Thelicensee(TVO)isupdatingcontainmentdeformationmeasurementsystemwithTe‐lemacpendulum type ofmeasuringdevices,which canmeasure also global horizontalandverticalmovementinthreelevels.Testpendulumswereinstalledyear2017inOL2before the outage. Measured deformations during leak tightness tests can be used toconfirmthebehaviourofthecontainmentinmorepreciseway.

Non‐DestructiveExamination(NDE)methodstogetherwiththedestructivetestsareal‐so needed in the future to confirm containment condition. The licensee (TVO) has anAMPfordestructiveandNDEtestsoftheconcretestructuresofthecontainment.(TVOGuide104756).ThepurposeoftheAMPistofindouttheeffectofagingtotheproper‐tiesoftheconcretestructuresofthecontainment.Anotherpurposeoftheprogrammeistoconfirmthattheprestressedstructureisbehavingasdesigned.TheearlierAMPofthelicensee(0‐TC‐O‐10/97)waspurelybasedonthetestingofcoredrills.Newprogrammewasstarted2005‐2006,whenpreliminaryNDEtestswithdifferentmethodswerecon‐ducted2005anddestructivetestswereconductedduring2006.Basedontheresultsofthe tests, the new programme (104756)was drafted. TheNDE research has been re‐peatedfirsttime2010.

TVOparticipatestheresearchprogrammeSAFIR2018andtheWandaproject,whichisconcentratedinconcreteNDEmethodsinvestigations.

TVOalsoparticipatesinEnergiforskresearchprogrammestogetherwithSwedishutili‐ties.

NPPunitOlkiluoto3FollowingdescriptionisbasedonAREVAreportsPECC‐G/2014/en/1023“In‐serviceIn‐spection Plan Civil Structures, 30UJA Building – Inner structures” andNGPM2/2004/en/0236 “In‐service Inspection Plan Civil Structures Containment”. TheguidesofthelicenseeTVOforOL3arestillunderpreparationandwillprobablyunifiedwiththeguidesforOL1andOL2.

ThepreviouslymentionedIn‐serviceInspection(ISI)planscomprisesprincipalrulesforperiodictestingandinspectionofthecontainmentbuildingstructuresoftheOlkiluoto3PowerPlantunit.

Containmentstructuralbehaviourunderpressureisstudiedintwotypeoftests;intheinitialstructuralintegritytest(ISIT)andintheISIleakagetest.TheISITduringcommis‐sioningofOL3nuclearpowerplanthasbeenconducted.

BeforetheISIT‐testfirstoverallleakagetest(typeA)wasconducted.Intheleakagetestthepressurewasincreaseduptothedesignpressure(5.3barabsolute).Acceptancecri‐



teriaoftheleakagerateforOL3was0.5Vol‐%per24hours.Theinitialstructuralinteg‐ritytestISITwasperformedafterthedeterminationoftheleakagerate.Aftertheinitialstructural integritytest(onehour in1,15timesthecontainmentdesignpressure) leaktightnesstestwasrepeatedsecondtime.

Followingmonitoringandinspectionactivitieshavebeenperformedfor innercontain‐mentstructureinconnectionofISIT:

Measuring of local strains in concrete by means of embedded acoustic straingaugesC110.Intheareaofthecylindricalshellthemeasuringpointsarelocatedat3differentverticalaxis(atopeningangleof67°,204°,294°)at5differentalti‐tudes(‐2.3m,+2.6m,+10.0m,+23.0m,+38.0m).Inthedomepartthemeasur‐ing points are located at 3 opening angles (67°, 204°, 294°), each at 2 levels+50.91m,+56.43mandinthecentreofthedome.Furthermoreinthebasematandareasaroundthelargerpenetrationsadditionalstraingaugeshavebeenim‐plemented.

Temperature measurements for the calibration of the strain gauge measure‐ments.Ateachlocationofasetofthreestraingauges(radial,tangential,orthog‐onal to this plane) there is one temperatureprobe implemented for themeas‐urementoftemperatureinconcrete.

Theelasticdeformationresponseofthecontainmenthasbeenmonitoredinor‐dertoprovideabasisfortheassessmentoftheintegrityandfunctionalityofthecontainment structure. The global displacements of the containment shell aremeasuredwithdirectpendulums,placedin3differentaxes(atopeningangleof67°,204°,294°)andinstalledat3differentlevels(+10.0m,+23.0m,+38.0m).Inaddition,theglobaldisplacementsofthedomeareahavebeenmeasuredwithla‐ser.

Monitoringoftheprestressingforcebydynamometersinstalledattheupperendof the fourvertical tendonsspacedout regularlyover thecircumferenceof thecylindricalshell.

The moisture of the concrete shell is measured by three concrete elementsstoredintheannuluswithsimilarmoistureconditionsasvalidfortheinnercon‐tainmentcylindricalconcreteshell.

Monitoringofcontainmentlinerstrainswithasystemofopticalstraingauges.

Visualsurfaceinspectionsandcrackmappingoftheinnercontainmentconcreteshell.Crackbridgingtransducershavebeenusedformonitoringofcrackwidthdevelopmentduringthepressuretest.

Visualsurfaceexaminationsforthecontainmentlinerandpenetrations,accord‐ingtoASMESectionXI.

PeriodicalISIleakageratetestunderpressureof4.9barabsolutewillbeperformedeve‐ry 5 years of operation (respectively 3 times in 15 years).Most of themeasurementswhichwereperformedduringtheISITshallberepeatedduringISItogetvalidcompari‐



sonvaluesduetotheelasticbehaviourofthecompleteconcretestructure.Thisindicatesthestageofageingofthecontainment.

Besidesuseforthetestswithoverpressure,themonitoringsystemisregularlyusedtofollowlongtermpropertiesofcontainmentstructure.Themeasurementsduringandbe‐tweenthetestsshallsupplythenecessarydatatoverifythatthesafetymarginsprovidedinthedesignoftheconcretestructureshavenotbeenreducedasaresultofoperatingandenvironmentalconditions.

Theinspectionsofcontainmentinternalstructuresaremainlyvisual.Theregularinspec‐tionswillstartaftertheplantisinoperation.Theproposedintervals(varyingfrom1to5yearsdependingonthetarget)fortheconcretesurfaceinspectionsandotheritemsofinterestare listedinreportPECC‐G/2014/en/1023.InspectionsubjectaregiveninthefollowingTable.

Furthermore,severalinvestigationprogrammeshavebeenperformedforOL3concretematerial.Forexample,creepandshrinkage,strengthdevelopmentandE‐modulusoftheconcretemixused in thecontainmenthavebeen investigated.Furthersamples for the



possible testingduring service lifetimeof 60 yearshavebeen taken and stored in theprestressinggalleryandthereactorbuilding.Inaddition36speciallocationsfordrillingcorespecimenshavebeenmarkedonthedomestructure.

InspectionhistoryofNPPunitOL3

TheInitialStructuralIntegrityTest(ISIT)oftheinnercontainmentstructureofOlkiluoto3hasbeenconducted.Theteststartedafterasubpressuretestphaseon2014‐01‐31andendedon2014‐02‐14.Themaximuminnerpressurewas6barabsoluteor5barrelative.

Measurementswerepreformedbefore,duringandafterthepressuretest.Targetvaluesforpendulumdisplacementsandstrainsofacousticalstraingaugeswerecomputedus‐ingFE‐analysisbytheDetailDesignerofthecontainment.Besidesthetargetvalues,alsolowerandupperlimitshavebeendetermined.ThelimitsforthedisplacementsfollowedthecriteriagiveninASMESectionIII,Division2.Forthestrainsrespectivelimitsweredetermined.

Allmeasurementresultswerebasedonrespectivezeromeasurementsimmediatelybe‐fore thepressure test. Themeasured strains in the current cylindrical part and in thedomepartofthecontainmentwallwerewithintheupperandlowerlimitsaccordingtoASME for pressurization as well as for depressurization phase. In the vicinity of theopeningsthemeasuredstrainsandcalculatedstrainsdeviatedfromeachotherasmallamount.AlsothedisplacementsmeasuredbypendulumswereinthelimitsaccordingtoASMEandthebehaviourofdisplacementswaslinearbothforpressurizationaswellasfordepressurizationphase.

ItwasconcludedthattheelasticbehaviouraswellastheintegrityoftheInnerContain‐mentwasensured.

7.1.4 Preventiveandremedialactionsforconcretestructures

TheNARshoulddescribekeypreventiveandremedialactionsthathavebeenidentifiedforeachNARexample.

Descriptionshouldinclude:

Criteriafortakingactions;

Proceduresfortakingactions;

Descriptionoftheactionstobetaken.


Preventiveactions

Inspections of the neoprenebearings between the steel girders and circularwall sup‐porting the roof of the reactor buildingwere conducted and follow –up of the defor‐mationshasbeengoingonsince2016.



Research and inspection programme on radiation effects to the concrete (biologicalshield)isgoingonandspecimenshavebeendrilled.

Theintegrityofthepaintingsoftheinteriorstructuresofthecontainmenthasbeenas‐suredbyrepairandrenovationworks.DBA‐testsandadhesiontestshavebeenconduct‐edforspecimens.

Inthedomepartofthesteelcontainmenttherehavebeendamagesofthecoatingcausedbythewaterleakagesoftheroofstructureofthereactorbuilding.Thedamageswerelo‐calandwererepairedyear2011asanormalmaintenanceworkinordertopreventcor‐rosion.

Remedialactions

TherewerealargestudyoftheanchorplatesandthecapacityofdrilledanchorboltsinLoviisaNPPunits,alotoftheanchorshadtoberepaired.

Themostcritical spots in thesteel containmentare theT‐weldsof theanchorageringand themounting flangebolts. Steel containmentmounting flangebolts havebeen in‐spected regularlyaccording to inspectionprogrammeandno significant corrosionhasbeendiscovered.Outershellmountingflangeboltswillbereplacedonceduringcurrentlicensingperiodaccordingtoexistingreplacementprogramme.

Therehasoccurredseveralwater leakages through the roof structureof theLoviisa1reactorbuildingatrecentyears.Thewaterinsulationstructureoftheroofisacombina‐tion glass fibre and bitumen felts. There have occurred cracking and blistering of theglassfibrefelts.Remedialactionshavebeendesignedontheyear2017.Newwaterinsu‐lationlayerwillbeinstalledandinthelowerpartsoftheroofandallthewaterandheatinsulation layerswill be replacedwith new layers.Water drainage sumps on the roofwillberenewedandairtightnessoftheroofhatchwillbeassuredwithnewsealing


Remedialactionsduringconstructionperiod

Duringconstructionphaseof theOL1andOL2containmentssomeconstructionerrorshavehappenedandresultsof theerrors mayhaveaneffectontheagingof thestruc‐tures.Someofthebuildingnon‐conformancesneededspecialstudiesandlargeremedialworks.Suchnon‐conformancesandremedialactionswereforexample;

InthecylinderpartofthecontainmentofOlkiluoto1unittherewassomelongercracks thatwereclosed, thesecrackshavebeenmonitoredduringseveraldec‐ades.Someofthem,whendevelopingwiderthancriteria0,2mm,havebeenin‐jected.

Therewasa fire in thereactorbuildingofOlkiluoto1unitduringconstructionphase8.2.1976.Firecauseddamagetotheupperpartofthecontainment,largeremedialworkswereneededandnewstructuralanalyseswereconducted.Re‐medialworksweredoneaccordingtorepairplansapprovedbySTUK.Accordingto the plans 1700 pcs of strengthening bolts (T25, A40, L=700) had to be in‐



stalledintheconcretingpart23,thethicknessofthewallofthecontainmenthadtobeincreasedduringtherepairworkofthecontainment.Therewerealsono‐ticedsomeconcretingerrorsneartheliner.Basedonthenewstructuralcalcula‐tionsdelivered to STUK itwas concluded that the loadbearing capacity of thestructureswasadequate.

Otherremedialactions

DisplacementmonitoringsystemofOlkiluoto1and2NPP´swasnotinitiallyexactlyac‐cordingtoASMESectionIIIDiv.2requirements.Measurementsystemdeviceswerein‐stallednearthemovementjointsnotinthepositionswhereASMErequires.Thereasonforthishasbeenpartlythefactthatthemovementjointswererathertightandtheyhadtobemonitored.Usingrelativedeformationsandmeasurementresultsofthemovementjointsithasbeenpossibletocontrolhowwellthemovingjointshavebeenfunctioning.Finally themovement jointswere so tight that theypreventedmovement. In order toguarantee the functionality, TVO finally had tomake a repair plan and enlargemove‐mentjointsbycuttingconcretearoundthejoints.

Deformationmeasurementswill be developed furtherwith Pendulum‐type of system.InstallationofthePendulum‐systemstarted2017.WhenthemeasurementresultsofthePendulum‐systemarecomparedtotheFE‐analysisresults,bettercomprehensionofthebehaviourofthecontainmentandpre‐stressingsystemcanbereached.Withthependu‐lums,thedeformationmeasurementsystemofOlkiluotounits1and2willsatisfythere‐quirementsofthestandardsASMEIIIDiv.2.andASMEXI.

DBA‐testswere conducted for thepaint specimensdrilledof the concretewallsof theOL1 andOL2 containments on the year 2009. Testswerepart of a TBY researchpro‐gramme, inwhichcommonrequirementsandguidelinesweredeveloped forpaintingsinside containments in Nordic countries. Painting specimens of OL1 fulfilled the re‐quirementsbutspecimensofOL2didnot.AsaresultofthestudyinnersurfacesoftheOL1andOL2containmentshavebeentotallyrepainted2010‐2011withapaintingsys‐temthatfulfilledtheDBA‐testsandthenewpaintingprogrammeAMP137614oftheli‐censee.

There have been several leakage cases of condensation pool (wet‐well) ofOlkiluoto 1unitontheyear1996,2003and2006.Latestleakageofthesteellinerofthecondensa‐tionpool(wet‐well)atOlkiluoto1wasrepairedontheyear2006andjointsoftheupperboundariesofthelinerofwet‐wellpoolatthebothunitsOlkiluoto1and2werealsore‐paired.

NPPunitOlkiluoto3

DuringconstructionphaseoftheOL3containmentandreactorbuildingsomeconstruc‐tionerrorshavehappenedandresultsoftheerrorsmayhaveaneffectontheagingofthestructures.Followingnon‐conformancesandremedialactionswerereportedduringconstruction;

During construction on the year 2008 there occurred a fire in the annulus be‐tween innerandoutershellof thereactorbuilding.Thedamagecausedby thefirewasactually ratherminorandstructureswererepaired.During leak tight‐



nesstests(ISI‐tests)theareaofthe firedamageisaspecialtargetofthevisualinspection.

Duringconstruction,crackswithconsiderablecrackwidthsoccurredatthereac‐torpitbetweenelevations+1,50mand+4,80duetohydrationheat.Tempera‐ture of the concreted area had been measured by temperature sensors. Non‐conformancereportandrepairplanwaswritten.Crackswererepairedbyinjec‐tion.InjectedcracksareaspeciallymarkedtargetofthevisualinspectionsoftheinnerstructuresoftheOlkiluoto3containmentduringLeak‐tightness(ISI)tests.Incaseofoccurringnewcracksinthesemarkedareasorinthevicinity,theloca‐tions,thecourses,thelengthsandthewidthsshallbedocumented.Alldetectedcracksatthisspecialareawithwidthsmorethan0,3mmshallbeinjected.

Duetothefabricationtechniqueoftheliner(largeprefabricatedblocks)ofOlki‐luoto3unittheinitialimperfectionswereratherlargeinthefinalproduct.Therewerealsosomeweldingerrorsdoneatthesite,thatneededtoberepaired.Theinitial imperfections increase the buckling stresses in the linerduringpossibleLOCA. That iswhy STUK demanded that the liner of the containment shall beequippedwithmonitoringdevicesi.e.additionalstrainmeasurementsystem.

7.2 Licensee’sexperienceoftheapplicationofAMPsforconcretestructures


Degradationmechanismsofsteelcontainmentandreactorbuildingconcretestructureshavebeenwellknownsincethestartofplantoperation.

Reactorbuilding concrete structureshavebeen inspected andmonitored according toprocedures and practices mentioned in Section 7.1.3 since 1978. Concrete structuresimportant to safety are cast‐in‐place and there are no post‐tensioned reinforced con‐cretestructuresinLoviisaplant.SupportingstructuresofRPVhavebeeninspectedfromconcretematerialsamplestakenfrombothunitswheninspectedcontainmentsteellinercondition.Basedontheinspectionsbothconcreteandlinerwereingoodcondition.

Noageing related severedegradationhavebeendetected in thecontainment concretestructures,steelcontainmentorpoollinersduringtheplantoperationhistory.

Floorcoatingsareingoodconditioninreactorbuildingandexpansionjointshavebeenrefurbishedbasedontheinspectionresults.

No changes have been done to the ageing management programme for containmentstructuresbasedontheinternalorexternaloperatingexperiences.

Steelcontainmentmountingflangeboltshavebeeninspectedregularlyaccordingtoin‐spection programme and no significant corrosion has been discovered. Outer shellmountingflangeboltswillbereplacedonceduringcurrentlicensingperiodaccordingtoexistingre‐placementprogramme.



Steelcontainmentandreactorbuildingconcretestructuresperiodic inspections,moni‐toring and testing is fairly comprehensive and unexpected degradation has not beenidentified.


OL1andOL2concretecontainmentin‐serviceinspections(ISI)havebeenrunningdur‐ingunitsoperation.AsaresultoftheISIprogrammesstatusofthecontainmentiscon‐trolledandISI‐programmesareinawaycontainmentsystem“health”programmes.

AMP104730Containmentdeformationmeasurementbetweencontainmentandreactorbuilding

DeformationmeasurementsbetweencontainmentandreactorbuildinginOL1andOL2startedthemiddleof1990’s.MeasurementsystemhasbeenoperatingwellandTVOhascollected20yearsinformation.

Thereasonwhymeasurementsystemdeviceswereinstallednearthemovementjointshas been the fact that the movement joints were rather tight. Using relative defor‐mationsandmeasurementresultsofthemovementjointsithasbeenpossibletocontrolhowwellthemovingjointshavebeenfunctioning.Inordertoguaranteethefunctionali‐tyofthemovementjoints,TVOfinallyhadtomakearepairplanandenlargemovementjointsbycuttingconcretearoundthejoints.

Asanexampledeformationmeasurementsduringcontainmentleaktightnesstest2016inOL1arepresentedintheFigures7.1below.TVOhasdevelopedtogetherEnergiforskFE‐modelofthecontainment.Verticalmeasurementresults(Figure7.1)arealmostthesameasthecalculatedresults(differencebelow10%).



Figure7.1.Measuredverticaldisplacementof the fuelpools(blue line)andcalculationresults(redsquares)duringcontainmentleaktightnesstest2016inOL1.

AMP104733Containmentconcretecrackmapping

Concretecrackrecordingandmapping(width,length,insomecasealsodepth)hasbeenperformedeveryyear.Thedevelopmentofcracksisfollowedinordertofindoutifthewidthorlengthofthecrackisincreasingorstayingconstant.Whenneededdevelopingcracksareinjected.CrackmappinghasbeenconductedaccordingtoASMESectionXI.

AMP104756Containmentconcretematerialfollow‐up

ThisAMPconcretematerialfollow‐upwasstartedwithtestsamplesandcontinuedwithdrillingcoresandNDEtests.TheNDEinvestigationshaveinvolvedfollowingmethods:

•IE–ImpactEcho

•SASW–SpectralAnalysisofSurfaceWaves



•MASW–MultichannelAnalysisofSurfaceWaves

AMP111409Containmentconcretehumidityfollow‐up

Concretehumiditymeasurementstarted20yearsagoanditseemsthatnochangeshashappened.Humiditysensorsmonitoralsoleakagesinthepoolsandgivesreliableinfor‐mationfortheconcretehumidityandshrinkage.Anexampleoftemperatureandhumid‐ityfollow‐upresultsduring2016inOL1isshowninFigure7.2below.

Figure7.2.Temperatureandhumidityduring2016inOL1.

AMP111556Containmentprestressingsteelfollow‐up

InOL1andOL2prestressedtestbeamshavebeenusedtostudythebehaviourandgiveinformationoftheprestressingtendonforcesandcorrosionrate.

InOL2pendulum typeofdisplacementmeasurementdeviceswillbe installed startingyear2017.WiththependulummeasurementscontainmentglobaldeformationscanberecordedinX‐,Y‐andZ‐directionsduringleaktightnesstestsandinthelongrun.Pendu‐lum‐system can be used also in the prediction prestressing tendon forces. In order tosucceedinthistask,resultsofanaccurateFE‐modelofthecontainmentareneeded.

Accordingtotheresultsthependulummeasurementsareratherthesameastheresultsof the measurement clocks and FE‐ analysis. An example of results showing defor‐mationsduringpressuretestinOL2(2017)isshownbelow(Fig.7.3).



Figure7.3HorizontaldisplacementinthemiddleheightofthecylinderwalloftheOL2containment.measuredbypenduluminstalled2017.

AMP113016Containmentconcretestraingauges

Straingaugesare installed to containment reinforcement in26positions. Straingaugemeasurementsaredoneduringnormaloperationandpressure/leaktightnesstests.AnexampleofthestrainmeasurementresultsduringleaktightnesstestisshowninFigure7.4below.Itcanbeseenthatthestrainsinrebarsarereversibleandelastic.

Figure7.4Strainintheuppergussetinthecorneroftheroofandcircularwallasafunc‐tionofpressure.LeaktightnesstestofOL2year2017.



NPPunitOlkiluoto3

Someexperiences from themost relevantAMPs forOL3 inner containmenthavebeencollectedalthoughthelifecycleoftheplantisonlyinthebeginning.

Monitoringofthelongtermbehaviouroftheprestressedcontainmentstructure:

Asstatedabove,withrespecttotheloadbearingfunctionoftheinnercontainmentcon‐cretewallandtheleaktightnessfunctionofthelinertheleadingageingmechanismsarecreep and shrinkage of the concrete in connection with the relaxation of the pre‐stressingsteel.Thecreepandshrinkagetogetherwiththerelaxationleadtoadecreaseofthepre‐stressingforceandadecreaseoftheconcretecompressionstress.Ifthistimedependent process would last unlimited it would finally cause tensile stresses andcracksintheconcreteundercertaindesignloadconditions.

Therefore the long termmonitoring of the loss of concrete compression stress canbeseenasoneofthemostrelevantAMPforOL3containment.Themonitoringisdonebystraingaugesandtemperatureprobesembeddedintheconcrete.Additionalinformationfrom relaxation of prestressing steel can be received through dynamometermeasure‐ments. Themonitoring has been started at the time of post‐tensioning in 2010 and itwouldbecontinueduntiltheendoflifetime.

Themeasurementresultsareregularlycomparedtotheoreticalcurveswhichhavebeencalculatedaccordingtotheformulasandassumptionsusedinthedetaileddesigncalcu‐lations of the containment, see example in the Figure below. Further investigationand/or assessmentwouldbenecessary if themeasured losseswould risehigher thanthecalculatedones.

Sofarthemeasuredstrainshavestayedbelowthecalculatedvalues.Thedevelopmentsofthemeasuredconcretestrainsalsoindicatethatthelongtermbehaviourofthepre‐stressedcontainmentstructurecorrelateswelltothecalculations.AnotherrelevantAMPapplicationisthemonitoringofthebehaviouroftheprestressedcontainmentstructureduringthetestswithoverpressureloading(ISIT,periodicalISI).Themonitoringisdonebystraingaugesandtemperatureprobesembeddedinthecon‐crete(concretestrains)aswellasbypendulums(displacements)andbyotherinstru‐mentssuchasdynamometers(tendonforce)andopticalstraingauges(linerstrains)



Figure7.5 Measuredtangentialstrainsatdifferentlevelsinthecylindricalwallcom‐

paredtopre‐calculatedconcretestrainlosses(greencurveaccordingtoassumptionsusedinthedesign).

Asstatedabove,theISITwasperformedin2014andthemonitoringresultsconfirmedtheelasticbehaviourandtheintegrityoftheinnercontainment.SimilarinvestigationswillberepeatedinthefutureinconnectionofperiodicalISItests.

7.3 Regulator’sassessmentandconclusionsonageingmanagementofconcretestructures

Regulator´sassessmentoftheageingmanagementprocessesSTUKguide[STUK‐GuideYVLE.6]requiresthatalicenseehasaplanforthemonitoringoftheageingdeformations,temperature,humidityoftheconcretestructuresofthecon‐tainment.Thein‐serviceinspectionplanshallpresenttheinspectionstobeconductedonstructuresatspecifiedintervalsduringplantoperation,themannerofperformanceoftheinspections,andthecriteriaforassessmentandrecordingoftheinspectionre‐sults.Theplanforthein‐serviceinspectionofreactorcontainmentconcretestructuresshallincludethefollowinginformation:

Inspectionofdisplacements,strainsandleak‐tightnessofstructuresatspecifiedintervalsandinconjunctionwithleakageandpressuretests.

Inspectionoftheconditionofpost‐tensionedcontainmenttendonsandanchor‐agesatspecifiedintervals.



Inspectionofstructuresessentialforthecontainment’sfunctionbytest loadingorotherreliablemethods,ifnecessary.

STUKguideE.6alsohasrequirementsofperiodicalinspectionsaccordingtoAMPs.Oth‐erwise ageing of concrete structures is managed by the general ageingmanagementprogrammes according to the STUK guide [STUK‐Guide YVL A.8]. STUK has inspectedandapproved

theageingmanagementprogrammes,

in‐serviceinspectionandmonitoringplansofcontainments,

leak‐tightnessandpressuretestprogrammesofthecontainments.

The ageingmanagement processes described in the ageingmanagement programmeshave been found satisfactory. In some cases STUK has required changes in the leaktightnessorpressuretestprogrammesorISI‐plansofcontainments.

For example STUK has required that the containment of Olkiluoto 3 NPP has to beequipped with strain measurement system of the steel liner. The reason for this re‐quirementwas large initial imperfectionscausedbythe largemodule fabricationtech‐nique.AfterrequiredchangestotheprogrammesSTUKapprovedtheISI‐plan.

Implementationoftheageingmanagementprogrammesandperiodicalinspectionshasbeen verified during the STUK inspections. The ageingmanagement process includesthat the licensee regularly evaluates the adequacy of containment ageing monitoringplanandthecoverageandeffectivenessofthewholeageingmanagementprogramme.Agoodexampleofthisisthatinstallationofpendulum‐typeofdeformationmeasurementsystemhasstartedinthecontainmentsofOL2andOL1ontheyear2017.

Regulator´s experience from inspection and assessment as part of its regulatory over‐sight

STUK reviews the annual ageingmanagement follow‐up reports of the licensees pro‐ducedaccordingtoGuideYVLA.8.Resultsofthecontainmentageingmonitoringshallbepresentedatleasteverythreeyearsinconnectionofthefollow‐upreport.

STUKinspectorsmakeobservationsabouttheageingoftheconcretestructuresincon‐nection of inspections during operation and outages of the NPP units, and during therenovationworks.STUKisdoinginspectionsatleastonceayearduringoutagesforeachNPPunitandonceayearamore thorough inspectioneitheroncivil structuresor fireprotection(KTO‐programmeofSTUK).

STUKreviewsfollowingreportsthataresenttoSTUKforinformationaftertestsandli‐censee´sinspections;

leak‐tightnessandpressuretestresults,

resultsofperiodicalinspections.



STUKalsoapprovestheconstructionplansofthemajorrenovationworksofsafetyclas‐sifiedconcretestructuresoftheNPP´s.UsuallySTUKhasmadesomerequirementsorremarksontheconstructionplanandpossiblydemandedsomeadditionalinformation.Licenseehasmadeanewrevisionwithcorrectivechangesintheplanandhasapprovedthenewversion.ThemainstrengthsandweaknessesthathavebeenidentifiedeitherbythelicenseesortheregulatorontheeffectivenessoftheSSCspecificAMPsTheperiodical test and inspectionprogrammesand In‐service inspectionprogrammeshaveturnedouttobeeffectivesothatnosignificantdegradationduetoageinghasbeenreported.

Duringconstructionphaseofthecontainmentsofthedifferentunitssomeconstructionerrorshavehappenedandresultsoftheerrorsmayhaveaneffectontheagingofthestructures.Someofthebuildingnon‐conformancesneededspecialstudiesandlargere‐medialworks(seeSection7.1.4).ContainmentsofOlkiluoto1and2arefunctioningintheleak‐tightnessandpressuretestsaccordingtolineartheoryanddeformationshavebeenreversibleinthelimitsofASMEIIIDiv.2..Themeasuredstrainsanddeformationsofthemonitoringsystemshavebeeninagoodcongruencewiththepre‐calculatedFE‐analysisresults.SofarthedisplacementmeasurementsystemofOlkiluoto1and2hasnotfulfilledtherequirementsofASMESectionIIIDiv.2.DeformationmeasurementswillbedevelopedfurtherwithPendulum‐typeofsystem.Withthependulumsthedeformationmeasure‐mentsystemofOlkiluoto1and2unitswillsatisfytherequirementsofthestandardsASMEIIIDiv.2.andASMEXI.SteelcontainmentofLoviisaunitsarefunctioninginalinearwayinleaktightnesstests.ImportantmountingflangeboltshavebeeninspectedregularlyaccordingtoinspectionprogrammeandnosignificantcorrosionhasbeendiscoveredinLoviisa1and2.Acer‐tainamountofboltsarechangedeveryyearinordertobesureofthecorrectbehaviour.ConcerningOlkiluoto3containmentmeasureddisplacementsandstrainshavebeenlin‐earandreversibleinthelimitsofASMEIIIDiv.2.STUKconsidersthattheageingmanagementprogrammesconcerningtheconcretecon‐tainmentsandreactorbuildingsoftheFinnishNPPunits,Loviisa1and2,Olkiluoto1,2and3,havebeenadequate.

8 Pre‐stressedconcretepressurevessels(AGR)

N/A



9 Overallassessmentandgeneralconclusions

9.1 Overallageingmanagement

Anewregulatoryguideforageingmanagement,i.e.,GuideYVLA.8hasbeenrecentlyis‐sued and enforced simultaneously with an overall revision of all STUK’s regulatoryguidesintheendof2013.InthisGuideYVLA.8thereisarequirementforthelicenseestodrawupanageingmanagementprogrammeforregulator’sapproval.SimilarbutlessdetailedrequirementswerestipulatedalsointhepreviousYVLguides.

Inthebeginningof2017TVOandFortumissuedtheirupdatedageingmanagementpro‐grammesac‐cordingtotheGuideYVLA.8.AfterareviewoftheprogrammesSTUKcon‐cluded that theboth ageingmanagementprogrammes shouldbedeveloped further tomeetthenewregulations.DeadlinefortherevisedprogrammeswassettotheendApril2018. Regardless of update needs of ageingmanagement programmes both licenseeshavecarriedoutreasonableageingmanagementfortheirSSCssincethecommissioningoftheplantunits.However,theroleofcomprehensiveageingmanagementprogrammeswillnowbemoreemphasizedastheoperationoftheNPPunitsisextendedbeyondtheoriginaldesignlifetime.EveninthiscasethedesignbasisoperabilityofSSCshastobemaintainedalthoughdegradationofSSCsmaybemoredifficulttoanticipateandcontrolinthelongtermoperation.

AgenericlessonlearnedinFinlandisthatthelicenseesshouldmakedecisionstomod‐ernizetheNPPsinduetimebeforetheendoftheirdesignlifetime.ApostponeddecisiontorenewforinstanceanI&Csystemoranelectricalsystemmaychallengepurchaseofnewsparepartsforthesystemsinservice.Thismayleadtosituationswherethelicen‐seemaynotbeabletodemonstratethesafetyofcontinuedoperationstotheregulator,oratleastthelicenseeandregulatormayunderstandtheadequatescopeofdemonstra‐tion inadifferentway.FinlandhassuccessfullyappliedPeriodicSafetyReviews(PSR)for the operatingNPPs. In the Finnish practice the licensee is obliged to demonstratethatthesafetyof theoperationscanbeensuredandimprovedalsoduringthenext10years.Thelicenseehastocommittosafetyimprovementsintermsofplantmoderniza‐tionstoaddressbothphysicalandtechnologicalageingofSSCs,too.

9.2 Electricalcables

Traditionally,sincemiddleof1980’s,STUK’sregulatoryguideshaverequiredfollow‐upprogrammes for thoseelectricalcables in thecontainment thatareneeded inaccidentsituations.Alltheothercablesimportanttosafetyarealsoincludedintheageingman‐agement programme for electrical equipment as stipulated in the earlier regulatoryguidesforelectricalequipment.Today,theregulatoryrequirementsfortheageingman‐agementofcablesimportanttosafetyarepresentedintheGuideYVLA.8andGuideYVLE.7.

STUKhasverifiedbyvariouskindof inspections that the licensee’s cableageingman‐agementprogrammesareinaccordancewiththeregulatoryrequirementswithoutma‐jordeviations.Atthemoment(endof2017),theageingmanagementprogrammeofOlk‐iluoto3isunderpreparation.



Theessentialcableageingmechanismsidentifiedbythelicenseesareambientheatload‐ing,conductorcurrentthermalloading,voltageload,ionizingandUVradiationloading,humidityandmechanicalstresses.Thecableenvironmentalmonitoringcomprisesusu‐allyof theambient temperaturesand radiation.Licensees’basic inspectionsarevisualand tactile inspections of the cables. Those inspections can also be executed for cablejoints, terminal boxes and cable trays specially in the containment.Heat loading is in‐spectedbymeasurementsandthermalimaging.Insomecasestandeltaandpartialdis‐charge testsareexecuted forMV‐cables.Samples fromcontainmentcabledepositsaretestedusingsuchcriteriaaselongationatbreak,tensilestrengthatbreak,andinsulationresistance.

Asremedialandpreventiveactions,minordefectsofjacketshavebeentemporarilyre‐paired.Iftheinspectionsorsampletestresultsindicatethatqualifiedlifeofthecableisneartheend,thenanewsimilarcableorqualifiednewcabletypehasbeeninstalledor,ifpossible,thequalifiedageofcableisextendedbyrequalification.

Incertaincasesthethermalorradiationloadofacableisreducedbychangingthecableroutes, by improving ventilationand coolingandby installing shielding structuresbe‐tweenthecableandradiationsource.

For instancestarting from1980’sLoviisaNPPhaveexecutedrelatively largeextensiveinspectionsandremedialactionsforelectricalcables.ActionsincluderequalificationofSiemensLOCAcables,improvingsteamgeneratorroomventilationandcooling,rerout‐ingandshieldingofcablesandreplacingdegradedcableswithnewones.

At NPP units Olkiluoto 1 and 2 electrical failures of in‐containment cables have beenveryrare.However,aspartofcontainmentcomponentrenewalsprojectELMAalsore‐newalsofin‐containmentcableswithlong‐termLOCA‐requirementshavebeenongoingsince2011.

As a summary the cable ageing management at the Finnish NPPs is currently imple‐mentedonsucha level that failurescausedbycabledegradationcausedbyageingarereportedonlyinfewcaseswerethedesigntemperatureofacable isexceededunfore‐seeably.Mostoftencablechangeshavebeenresultsofchangesintheequipmenttypesordue to themechanicaldamages causedby theworksnear cabling. In somecases itwasreportedthatacabletypehadbecomeobsoletebecauselackofrepairpartsandin‐stallationexpertise.

9.3 Reactorpressurevessels

STUKhas inspectedLicensees’ ageingmanagementprogrammes includingRPV ageingmonitoringplans.Implementationoftheageingmanagementprogrammeshasbeenver‐ifiedduringSTUKinspections.Theageingmanagementprocessesdescribedintheage‐ingmanagementprogrammeswerefoundsatisfactoryattheoperatingNPPs.TheageingmanagementprogrammesofOlkiluoto3arepracticallycompletedalthoughtheunit isstillundercommissioning.

STUK considers that the ageingmanagement programmes concerning theRPVs of theFinnishNPPunitsLoviisa1,Loviisa2,Olkiluoto1,Olkiluoto2andOlkiluoto3areade‐quate.



One significant ageingmanagement issue related to RPVs is the irradiation embrittle‐mentofLoviisaRPVsandthethermalannealingofthecoreareaweldofLoviisa1RPVin1996.The embrittlement rateof the critical coreareaweldshas tobe carefullymoni‐toredbythesurveillanceprogrammesaslongastheRPVsareinoperation.Ifthelicen‐seeplanstooperatetheplantunitsbeyondthecurrentoperatinglicenses,i.e.,beyond50years of operation, somemeasuresmaybenecessary to confirm safe operationof theRPVs.

Sofaroneindicationhasbeendetectedinalowpressuresafetyinjection(TH)nozzleofLoviisa1RPV.Itmaybecomeanageingmanagementissueifnewindicationswillbede‐tected inothernozzlesof samekind in future inspections.However, it is alsopossiblethattheexistingindicationprovesouttobeamanufacturingdefect.

Thedetectedindicationsinthenozzle/safe‐endweldsofsystems312(feedwater)and323 (reactor core spray) may become a significant ageingmanagement issue at NPPunitsOlkiluoto1and2.ThelicenseeTVOhasdecidedtocontinueitsworkandinspectand, ifnecessary,repair thecorrespondingweldsofsystems312,323andalsosystem321(shutdowncooling)atbothunitsOL1andOL2.

TheoperatinglicenserenewalprocessforNPPunitsOlkiluoto1and2iscurrentlyunderway.ThelicenseeTVOhasappliedpermissiontocontinueoperatingNPPunitsOL1andOL2until60years.InthiscontextTVOhasupdatedthestrengthandfatigueanalysesoftheRPVs. These analyseshavebeenquite comprehensive including almost everythingnecessaryexcept forbrittleand fast fractureanalyses thatarerequired inSection6of[STUK–GuideYVLE.4].TVOhasbeengiventimetocompletetheanalyses.Allrelevantageingmechanismshavetobeconsideredintheanalyses.

Theanalysesperformedso farshowthat themostcriticalsubcomponentsof theRPVsarethefeedwater(312)nozzles.Basedontheseanalyses,thedamageriskofallotherRPV’s in‐scope subcomponents ismuch lower.Theanalyses also show that IGSCCandIASCCcanaccelerateageingoftheRPVs.

9.4 Concretecontainmentstructures

STUK reviews the annual ageingmanagement follow‐up reports of the licensees pro‐duced according toGuideYVLA.8. The ageingmanagement of buildings shall be pre‐sentedatleasteverythreeyearsinconnectionofthefollow‐upreport.STUKisalsocar‐ryingoutinspectionsconcerningtheageingoftheconcretestructuresduringplantop‐erationandoutages.

Licenseesarecarryingout leak‐tightnessandpressuretestsandperiodical inspectionsforthecontainment.ResultreportsofthesearesenttoSTUKforreview.

DuringconstructionphaseofthecontainmentsofNPPunitsLoviisa1and2andOlkiluo‐tounits1and3someconstructionerrorsandaccidentshavehappenedwhichmayhaveaneffectontheagingofthestructures.Someofthebuildingnon‐conformancesneededspecialstudiesandlargeremedialworks,whichweresuccessful(seeSection7.1.4).

ContainmentsofOlkiluotounits1and2arefunctioninginthe leak‐tightnessandpres‐suretestsaccordingtolineartheoryanddeformationshavebeenreversibleinthelimits



ofASME IIIDiv. 2. Themeasured strains anddeformationsof themonitoring systemshavebeeninagoodcongruencewiththepre‐calculatedFE‐analysisresults.

So far thedisplacementmeasurementsystemofOlkiluoto1and2hasnot fulfilled therequirementsofASMESectionIIIDiv.2.DeformationmeasurementswillbedevelopedfurtherwithPendulum‐typeofsystem.WhenthemeasurementresultsofthePendulum‐systemarecomparedtotheFE‐analysisresults,bettercomprehensionofthebehaviourof thecontainmentandpre‐stressingsystemcanbereached.With thependulums, thedeformationmeasurement system of Olkiluoto units 1 and 2 will satisfy the require‐mentsofthestandardsASMEIIIDiv.2.andASMEXI.

SteelcontainmentsofLoviisaunitsarefunctioninginalinearwayinleak‐tightnesstests.The licenseehas inspected importantmounting flangebolts regularlyaccording to in‐spectionprogramandnosignificantcorrosionhasbeendiscovered.Acertainamountofboltsarechangedeveryyearinordertobesureofthecorrectbehaviour.

ConcerningtheOlkiluoto3containment,measureddisplacementsandstrainshavebeenlinearandreversibleinthelimitsofASMEIIIDiv.2duringthefirstpressuretest(InitialStructuralIntegrityTestISIT).

STUKconsidersthattheageingmanagementprogrammesconcerningtheconcretecon‐tainmentsandreactorbuildingsoftheFinnishNPPunitsareadequate.



10 References

ACI 349‐76, ACI Standard: Code Requirements for Nuclear Safety Related ConcreteStructures1976

ASMEBoiler&PressureVesselCode,SectionIII:RulesforConstructionofNuclearFacili‐tyComponents‐Division2‐CodeforConcreteContainments

ASMEBoiler&PressureVesselCode,SectionXI:Rules for In‐service InspectionofNu‐clearPowerPlantComponents

BY50,ConcreteCode2012,ConcreteAssociationofFinland

ChemistryHandbookPartVMonitoring,FGF,STC‐G/2008/en/052,Rev.B

Cronvall,O.SusceptibilityofBWRRPVandItsInternalstoDegradationMechanisms.Tobepublished2018.

Draft Regulatory Guide (DG)‐1197, Inservice Inspection of Prestressed Concrete Con‐tainment Structures with Grouted Tendons (proposed revision 2 of Regulatory Guide1.90,August1977).

FGFNTCM‐G/2009/en/1013rev.B,SpecialAgeingManagementProgram,ReactorPres‐sureVessel,ArevaNP

IAEASafetyofNuclearPowerPlants:Design;NoSSR‐2/1

IAEASafetyofNuclearPowerPlants:CommissioningandOperation;NoSSR‐2/2

IAEASafetyReportsSeriesNo.82,AgeingManagementforNuclearPowerPlants:Inter‐nationalGenericAgeingLessonsLearned(IGALL)

IAEASafetyReports SeriesNo.57, SafeLongTermOperationofNuclearPowerPlants,2008

IAEASafetyStandardsSeriesNo.NS‐G‐2.6,SafetyGuide,Maintenance,Surveil‐lanceandIn‐ServiceInspectioninNuclearPowerPlants,Vienna,2002

IAEASafetyStandardsSeriesNo.NS‐G‐2.12AgeingManagementforNuclearPowerPlants

IAEASafetyStandards,SpecificSafetyGuideNo.SSG‐13,ChemistryProgrammeforWa‐terCooledNuclearPowerPlants,

IAEANuclearEnergySeriesNo.NP‐T‐3.13,StressCorrosionCrackinginLightWaterRe‐actors:GoodPracticesandLessonsLearned

LO1‐K822‐00044versio2.0,Loviisa1ja2,Reaktoripainesäiliöidenikääntymisenhallin‐ta;yhteenveto.19.6.2014.



Nevander,O.OL1–Finalsafetyanalysisreportforsystem211–Reactorpressurevessel.FSAR,ReportNo.106076,18.6.2004.44p.

Nuclear Plant Chemistry Conference, Paper Reference n°167 046, The Application ofFilm‐formingAminesinsecondarysidechemistrytreatmentofNPPs,Paris2012

NUREG‐1801,Rev.2,GenericAgingLessonsLearned(GALL)Report(Volumes1and2),U.S.NuclearRegulatoryCommission,2010

OL1andOL2Document108654:“OL1jaOL2‐Suojarakennuksenkaapelienkunnonval‐vonta(tarkastuksetjakoestukset)”.

OL1, OL2, OL3 and KPA Document 117279, “OL1/OL2/OL3 ja KPA Ikääntymisenhallintaohjelma”

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RIL 53d Finnish Concrete Code 1971 in Finnish; Betoninormit 1971 SuomenRakennusinsinöörienLiitto.

RIL53d, FinnishConcreteCode1975 inFinnish;Betoninormit1975,Betonielementti‐normit1975,Rajatilamitoitusohjeet1975

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STUK–GuideYVL3.8:NuclearPowerPlantPressureVesselIn‐serviceInspectionwithNon‐DestructiveTestingMethods,22.09.2003

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STUK–GuideYVLE.4:Strengthanalysisofnuclearpowerplantpressureequipment,15November2013

STUK–GuideYVLE.5:In‐serviceinspectionofnuclearfacilitypressureequipmentwithnon‐destructivetestingmethods,20May2014

STUK–GuideYVLE.6:Buildingsandstructuresofanuclearfacility,15November2013

STUK–GuideYVLE.7:ElectricalandI&Cequipmentofanuclearfacility,15November2013

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11 ANNEX1:FinnishNPPsites.

Loviisa1(right)andLoviisa2(left).

Olkiluoto1(middle),Olkiluoto2(left)andOlkiluoto3.



12 ANNEX2:KeytechnicaldocumentationofNPPunitsOlkiluoto1and2cabling

129251(TE‐Be‐11e) GeneralTechnicalRequirementsfordesign,manufacturinganddeliveryofcablesforpowersupply,measuringandcontrolpur‐poses

107231(TE‐Be‐1e) TechnicalRequirementsconcerningtransportandstorageandnormaloperatingconditionsforelectricalandcontrolequipment

103394(TE‐Be‐2e) TechnicalRequirementsconcerningaccidentoperationcondi‐tionsforelectricalandcontrolequipment

129248(TE‐Ot‐130e) InspectionInstructionforDBEtestingofcables

102478(TE‐Ot‐135e) InspectionInstructionforradiationandthermalexposuretestofinsulatedwires

132743 Suitabilityanalysis,“OL1/OL2LOCA”cables,manufacturedbyHabiaCableAB

0‐2/1/151 Pre‐inspectiondocumentation,FirewallIIIcablesforin‐containmentuseinLOCA‐conditions,manufacturedbyTheRock‐bestosCompany

0‐2/531/26 Pre‐inspectiondocumentation,includingthetypetestdocumen‐tationforneutronfluxmeasurementcoaxialcables,manufacturedbyHabiaCableAB

135557 Typetestdocumentation,Lipaloncables,manufacturedbyLilje‐holmensKabelfabrikABandAseaKabelAB

168775 Suitabilityanalysis,halogenfreestandardcablesforoutsidecon‐tainmentuse,manufacturedbyRekaKaapeliOy

119125 Suitabilityanalysis,standardcablesforoutsidecontainmentuse,manufacturedbyRekaKaapeliOy

126857 Suitabilityanalysis,standardcablesforoutsidecontainmentuse,manufacturedbyDrakaNKCablesOy

0‐2/515/1 Pre‐inspectiondocumentation,standardinstrumentationcablesforoutsidecontainmentuse,manufacturedbyNokiaCablesLtd

168543 Lifetimeassessment,in‐containmentcables



13 ANNEX3:KeytechnicalOL3projectdocumentationrelevanttocabling

EZE‐100032 ProjectSpecification,normalpowerandcontrolcables

NLEE‐G/2006/en/1019 ProjectSpecification,normalI&Ccables

EYTM‐100252 ProjectSpecification,accidentproofedpowerandcon‐trolcables

EYTM‐100366 ProjectSpecificationforAccidentProofedI&C‐Cables<60VoftheNI

PELES‐G/2010/en/1001 TechnicalSpecification,specialaccidentproofedI&Ccables

EZE‐100050 EquipmentQualificationSpecification,normalpowerandcontrolcables

NLEE‐G/2008/en/1002 EquipmentQualificationSpecification,normalI&Cca‐bles

EYTM‐100430 EquipmentQualificationSpecification,accidentproofedpowerandI&Ccables

NLQ‐G/2007/en/1005 Suitabilityanalysis,normalpowerandcontrolcables

NLQ‐G/2007/en/1010 Suitabilityanalysis,accidentproofedpowerandcontrolcables

NLM‐G/2009/en/1001 Suitabilityanalysis,normalI&Ccables

NLM‐G/2009/en/1002 Suitabilityanalysis,accidentproofedI&Ccables,manu‐facturerLEONI

IBQ‐G/2011/en/1007 Suitabilityanalysis,I&Cextensioncables,manufacturerHEW

IBQ‐G/2011/en/1008 Suitabilityanalysis,specialaccidentproofedI&Ccables,manufacturerEUPEN

NLLN‐G/2008/en/1081 Suitabilityanalysis,ECIconnectingline

NLLN‐G/2008/en/1082 Suitabilityanalysis,RPVLconnectingline

IBQ‐G/2014/en/1001 Suitabilityanalysis,N16cables



14 ANNEX4:ASchematicdrawingsofLoviisa1andLoviisa2RPVs



15 ANNEX5:AschematicdrawingofOlkiluoto1andOlkiluoto2RPVs



16 ANNEX6:AschematicdrawingofOlkiluoto3RPV



17 ANNEX7 SummaryofOlkiluoto3RPVsubcomponents’materials

Itemno. Description Material/FillerMetal WeldProcess

1 LowerHead Low‐AlloySteel16MND5

Weld–LowerHeadtoTransitionRing

Low‐AlloySteelUS‐56B/MF‐27

SAW

CladdingLowerHeadtoTransitionRing

AusteniticStainlessSteelUSB‐309LK,USB‐308LK,PFB‐7FK,NC‐39LK,NC‐38LK

SMAW,SAW

CladdingLowerHead(CentralZone)

AusteniticStainlessSteelNC‐39LK,NC‐38LK

SMAW

2 TransitionRing Low‐AlloySteel16MND5

Weld–CoreShelltoTransitionRing

Low‐AlloySteelUS‐56B/MF‐27 SAW

CladdingCoreShellandTransitionRing


SMAW,SAW

Weld–RadialGuidetoTransitionRing(DMW)

Nickel‐base+Low‐AlloyWELAC152

SMAW

ButteringforRadialGuides Low‐AlloySteelWELAC152 SMAW

4 CoreShell Low‐AlloySteel16MND5

Weld–CoreShelltoIntegratedShell,Weld–CoreShell


CladdingCoreShellandIntegratedShell


SMAW,SAW

5 Nozzle/Flange‐integratedShell

Low‐AlloySteel16MND5

Weld–LeakDetectionTubetoIntegratedShell,Weld–SealLedgetoIntegratedShell

AusteniticStainlessSteelTGS‐308LK TIG

ButteringforLeakDetec‐tionTubeWeldtoInte‐gratedShell


SMAW



CladdingIntegratedShellkeyway,CladdingIntegratedShell


SMAW

CladdingIntegratedShell,CladdingIntegratedShellinternalssupportledge


SMAW,SAW

6,7 Inlet/OutletNozzle Low‐AlloySteel16MND5

Weld–Inlet/OutletNozzletoIntegratedShell


CladdingofInlet/OutletNozzleBore


SMAW,SAW

CladdingofInlet/OutletNozzleandIntegratedShell

AusteniticStainlessSteelTGS‐309LK,TGS‐308LK,NC‐39LK,NC‐38LK

SMAW,TIG


SMAW

8 Safe‐End AusteniticStainlessSteel(nitrogencontrolled)Z2CND18‐12

Weld–Inlet/OutletNozzletoSafe‐End(DMW)

Nickel‐baseAlloyWELTIG52

TIG

11 HeadFlange Low‐AlloySteel16MND5

Weld–HeadFlangetoUpperHead


CladdingHeadFlangeandUpperHead,CladdingHeadFlange


SMAW,SAW

CladdingHeadFlangekeyway


SMAW

12 UpperHead Low‐AlloySteel16MND5

Weld–HandlingLugtoUpperHead

Low‐AlloySteelLBL‐96 SMAW

CladdingUpperHead(CentralZone)


SMAW



18 ANNEX8:AschematicdrawingofLoviisa1and2reactorbuilding



19 ANNEX9:AschematicdrawingofOlkiluoto1and2containment

FigureSchematicdrawingpresentingOlkiluoto1and2cross‐sectionofthecontainment.



20 ANNEX10:AschematicdrawingofOlkiluoto3reactorbuilding

FigureSchematicdrawingpresentingOlkiluoto3.cross‐sectionofthereactorbuildingwithinnerandoutercontainment.IntheFigureboundariesbetweenSafetyClass2and3structureshasbeenpresented.

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