ML111360432: Difference between revisions

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{{Adams
#REDIRECT [[IR 05000443/2011007]]
| number = ML111360432
| issue date = 05/23/2011
| title = IR 05000443-11-007; March 7-11, 21-25, and April 4-8, 2011; Seabrook Station; Inspection of the Scoping of Non-Safety Systems and the Proposed Aging Management Procedures for the NextEra Energy Seabrook LLC Application for Renewed License
| author name = Conte R J
| author affiliation = NRC/RGN-I/DRS/EB1
| addressee name = Freeman P
| addressee affiliation = NextEra Energy Seabrook, LLC
| docket = 05000443
| license number = NPF-086
| contact person =
| document report number = IR-11-007
| document type = Inspection Report, Letter
| page count = 31
}}
See also: [[followed by::IR 05000443/2011007]]
 
=Text=
{{#Wiki_filter:UNITED STATESNUCLEAR REGU LATORY COMMISSIONREGION I475 ALLENDALE ROADKlNG OF PRUSSlA. PA 19406-1415l{ay 23, 20ILMr. Paul FreemanSite Vice PresidentNextEra Energy Seabrook LLCP. O. Box 300Seabrook, NH 03874SUBJECT: NEXTERA ENERGY SEABROOK - NRC LICENSE RENEWAL INSPECTIONREPORT 05000443/201 1 007Dear Mr.On April 8, 2011, the NRC completed the onsite portion of the inspection of your application forlicense renewal of Seabrook Station. The NRC inspection is one of several inputs into the NRCreview process for license renewal applications. The enclosed report documents the results ofthe inspection, which were discussed on March 28rh and April 8th with members of your staff.The purpose of this inspection was to examine the plant activities and documents that supportthe application for a renewed license of Seabrook Station. lnspectors reviewed the screeningand scoping of non-safety related systems, structures, and components, as required in10 CFR 54.4(a)(2), to determine if the proposed aging management programs are capable ofreasonably managing the effects of aging.The inspection team concluded screening and scoping of non-safety related systems,structures, and components, was implemented as required in 10 CFR 54.4(a)(2), and the agingmanagement portion of the license renewal activities were conducted as described in theLicense Renewal Application.We noted that your staff continued to develop an appropriate initial response to the aging effectof the alkali-silica reaction in certain concrete structures of Seabrook Station. Because yourinvestigation and testing was ongoing and you were not currently in a position to propose a newor revised aging management program, the inspection team was unable to arrive at aconclusion about the adequacy of your aging management review for the alkali-silica reactionissue. As part of the ongoing review of your application for a renewed license, you shouldcontinue to inform the Division of License Renewal as you develop your response to the alkali-silica reaction issue. With assistance from our Headquarters Office, Region I will review thosekey points in the implementation of your project plan associated with this issue to ensure thecurrent licensing bases is maintained, a key assumption in the license renewal process.Except for the alkali-silica reaction issue, the inspection results support a conclusion ofreasonable assurance with respect to managing the effects of aging in the systems, structures,and components identified in your application. The inspection also concluded thedocumentation supporting the application was in an auditable and retrievable form.
P. FreemanIn accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosure will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records (PARS) component of NRC's document system(ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).Sincerely,6LA.-/PetuRichard J. Conte, ChiefEngineering Branch 1Division of Reactor SafetyDocket No. 50-443License No. NPF-86Enclosure: Inspection Report0500044312011007cc Mencl: Distribution via ListServ
P. FreemanIn accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosure will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records (PARS) component of NRC's document system(ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.qovireadinq-rmladams.html (the Public Electronic Reading Room).Sincerely,/RNRichard J. Conte, ChiefEngineering Branch 1Division of Reactor SafetyDocket No. 50-443License No. NPF-86Enclosure:cc Mencl:Distribution Mencl: (VlA E-MAIL)W. Dean, RAD. Lew, DRAP. Wilson, DRSA. Burrit, DRPC. LaRegina, DRPI nspection Report 05000443/201 1 007Distribution via ListServA. Williams, Rl OEDOROPreports@nrc.govD. Bearde, DRSRegion I Docket Room (with concurrences)SUNSI Review Gomplete:MCM/RJC(Reviewer's lnitials)ADAMS ACC#MLI11360432DOCUMENT NAME: G:\DRS\Engineering Branch 1\_Technical lmportance\SeabrookConcrete\SbkLRl Rpts\05000443 201 1 007 lP7 1 OO2 Sbrk I nsp Rpt Final. docxAfter declaring this document "An Official Agency Record" it will be released to the Public.To receive a copy of this document, indicate in the box: 'C" = Copy without attachmenVenclosure "E" = Copy WthattachmenVenclosure "N" = Noost18t11OFFICIAL RECORD COPY
Docket No:License No:Report No:Licensee:Facility:Location:U. S. NUCLEAR REGULATORY COMMISSIONREGION I50-443NPF-8605000443/2011007NextEra Energy Seabrook LLCSeabrook StationSeabrook, NHMarch 7-11,21-25, and April 4-8,2011M. Modes, Team Leader, Division of Reactor Safety (DRS)G. Meyer, Sr. Reactor Inspector, DRSS. Chaudhary, Reactor Inspector, DRSJ. Lilliendahl, Reactor Inspector, DRSRichard J. Conte, ChiefEngineering Branch 1Division of Reactor Safety
SUMMARY OF FINDINGSlR 0500044312011007; March 7-11,21-25, and April 4-8,2011, Seabrook Station; Inspection ofthe Scoping of Non-Safety Systems and the Proposed Aging Management Procedures for theNextEra Energy Seabrook LLC Application for Renewed License for Seabrook Station.This inspection of license renewal activities was performed by four regional office engineeringinspectors. The inspection was conducted in accordance with NRC Manual Chapter 2516 andNRC lnspection Procedure 71002. This inspection did not identify any "findings" as defined inNRC Manual Chapter 0612. The inspection team concluded screening and scoping of non-safety related systems, structures, and components, were implemented as required in 10 CFR54.4(a)(2), and the aging management portions of the license renewal activities were conductedas described in the License RenewalApplication. Except for the alkali-silica reaction issue, theinspection results support a conclusion of reasonable assurance with respect to managing theeffects of aging in the systems, structures, and components identified in your application. Theinspection concluded the documentation supporting the application was in an auditable andretrievable form.
40.A21REPORT DETAILSOther - License RenewalInspection ScopeThis inspection was conducted by NRC Region I based inspectors in order to evaluatethe thoroughness and accuracy of the screening and scoping of non-safety relatedsystems, structures, and components, as required in 10 CFR 54.4(a)(2) and to evaluatewhether aging management programs will be capable of managing the identified agingeffect in a reasonable manner.The team selected a number of systems for review, using the NRC accepted guidance; inorder to determine if the methodology applied by the applicant appropriately captured thenon-safety systems affecting the safety functions of a system, component, or structurewithin the scope of license renewal.The team selected a sample of aging management programs to verify the adequacy ofthe applicant's documentation and implementation activities. The selected agingmanagement programs were reviewed to determine whether the proposed agingmanagement implementing process would adequately manage the effects of aging on thesystem.The team selected risk significant systems and conducted a review of the AgingManagement Basis documents for each selected system to determine if the applicant hadadequately applied the Aging Management Programs to ensure that reasonableassurance exists for the monitoring of aging effects on the selected systems.The team reviewed supporting documentation and interviewed applicant personnel toconfirm the accuracy of the license renewal application conclusions. For a sample ofplant systems and structures, the team performed visual examinations of accessibleportions of the systems to observe aging effects.Scopinq of Non Safetv-Related Svstems. Structures. and Components under10 CFR 54.4 (a) (2)For scoping the team reviewed program guidance procedures and summaries of scopingresults for Seabrook Station to assess the thoroughness and accuracy of the methodsused to bring systems, structures, and components within the scope of license renewalinto the application, including non-safety-related systems, structures, and components, asrequired in 10 CFR 54.4 (a)(2). The team determined that the procedures wereconsistent with the NRC accepted guidance in Sections 3, 4, and 5 of Appendix F toNuclear Energy Institute (NEl) 95-10, Rev. 6, "lndustry Guideline for lmplementing theRequirements of 10 CFR Part 54," (Section 3: non-safety-related systems, structures,and components within scope of the current licensing basis; Section 4: non-safety-relatedsystems, structures, and components directly connected to safety-related systems,structures, and components; and Section 5: non-safety-related systems, structures, andcomponents not directly connected to safety-related systems, structures, anda.Enclosure
2components). The team noted that scoping guidance was not clear regarding structuraldescriptions. By drawing reviews and in-plant walk-downs, the team identified that thefew scoping errors related to the guidance inconsistencies were conservative, i.e.,components were placed within the scope of license renewalwhich were not required tobe included. Subsequently, the applicant revised the scoping guidance, and the teamreviewed the revised guidance.The team reviewed the set of license renewal drawings submitted with the SeabrookStation License Renewal Application, which was color-coded to indicate non-safetyrelated systems and components in scope for license renewal. The drawings includednumerous explanatory notes, which described the basis for scoping determinations onthe drawings. The team interviewed personnel, reviewed license renewal programdocuments, and independently inspected numerous areas within Seabrook Station, toconfirm that appropriate non-safety-related systems, structures, and components hadbeen included within the license renewal scope; that systems, structures, andcomponents excluded from the license renewal scope had an acceptable basis; and thatthe boundary for determining license renewal scope within the systems, including seismicsupports and anchors, was appropriate.The Seabrook Station in-plant areas reviewed included the following:. Turbine Buildingo Primary Auxiliary Building. East Main Steam & Feedwater Pipe Chaseo West Main Steam & Feedwater Pipe Chase. Control Building. Service Water Pumphousee Emergency Feedwater Pumphouse and Pre-Action Valve Buildingo Steam Generator Blowdown Buildingo Emergency Diesel Generator Room B. RCATunnel. Tank Farm AreaFor systems, structures, and components selected regarding spatial interaction (failure ofnon-safety-related components adversely affecting adjacent safety-related components),the team determined the in-plant configuration was accurately and acceptablycategorized within the license renewal program documents. The team determined thepersonnel involved in the process were knowledgeable and appropriately trained.For systems, structures, and components selected regarding structural interaction(seismic design of safety-related components dependent upon non-safety-relatedcomponents), the team determined that structural boundaries had been accuratelydetermined and categorized within the license renewal program documents. The teamdetermined that the applicant had thoroughly reviewed applicable isometric drawings todetermine the seismic design boundaries and had correctly included the applicablecomponents in the license renewal application, based on the inspector's independentEnclosure
3review of a sample of the isometric drawings and the seismic boundary determinationscombined with in-plant review of the configurations.ln summary, the team concluded that the applicant had implemented an acceptablemethod of scoping of non-safety-related systems, structures, and components and thatthis method resulted in accurate scoping determinations.Proorams8.2.1.9 Bolting InteqritvThe Seabrook Station Bolting Integrity Program is an existing program that manages theaging effects of cracking due to stress corrosion cracking, loss of material due to general,crevice, pitting, and galvanic corrosion; Microbiologically induced corrosion, fouling andwear; and loss of preload due to thermal effects, gasket creep, and self-looseningassociated with bolted connections. The program manages these aging effects throughthe performance of periodic inspections. The program also includes repair/replacementcontrols for ASME Section Xl related bolting and generic guidance for material selection,thread lubrication and assembly of bolted joints.The inspector reviewed the program basis documents, program description, baselineinspection results, subsequent inspection results for trending, and implementingprocedures to determine the scope and technical adequacy of the Program. Also, theteam reviewed action requests (ARs) to assess the adequacy of evaluations of findings,and resolution of concerns, if any, identified in these inspections.The inspector noted that the program follows the guidelines and recommendationsprovided in NUREG-1339, "Resolution of Generic Safety lssue 29; Bolting Degradation orFailure of Bolting in Nuclear Power Plants", EPRI NP-5769, "Degradation and Failureof Bolting in Nuclear Power Plants" (with the exceptions noted in NUREG- 1339), andEPRI TR-104213, "Bolted Joint Maintenance and Application Guide" for comprehensivebolting maintenance. However, indications of aging identified in ASME pressure retainingbolting during In-service Inspection are evaluated per ASME Section Xl, Subsections3600. lndications of aging identified in other pressure retaining bolting, nuclear steamsupply system component supports, or structural bolting are evaluated through theCorrective Action Program,This program covers bolting within the scope of license renewal, including:1. safety- related bolting,2. bolting for nuclear steam supply system component supports,3. bolting for other pressure retaining components, including non-safety relatedbolting; and,4. structural bolting.The aging management of reactor head closure studs is addressed by Seabrook StationReactor Head Closure Studs Program (8.2.1.3) and is not part of this programlEnclosure
4B.2.1.13 lnspection of Overhead Heavy Load And Liqht Load (Related To Refuelinq)Handlino SvstemsThe Seabrook Station Inspection of Overhead Heavy Load and Light Load (Related toRefueling) Handling Systems Program is an existing program that will be enhanced tomanage the aging effects of loss of material due to general corrosion and due to wear ofstructural components of lifting systems and the effects of loss of material due to wear onthe rails in the rail system, for lifting systems within the scope of license renewal.Included in scope are those cranes encompassed by the Seabrook Station commitmentsto NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," plus two cranesrelated to fuel handling.The team reviewed the program basis documents, program description, baselineinspection results, subsequent inspection results for trending, and implementingprocedures to determine the scope and technical adequacy of the Program. Also, theteam reviewed ARs and work orders to assess the adequacy of evaluations of findings,and resolution of concerns, if any, identified in these inspections.The team noted that the program manages loss of material due to general corrosion onstructural steel members and rails of the cranes within the scope of license renewalincluding the structural steel members of the bridges, trolleys and monorails. Theprogram also manages loss of material due to wear on rails. Only the structural portionsof the in-scope cranes and monorails are in the scope of this program. The individualcomponents of these overhead handling systems that are subject to periodicreplacement, or those which perform their intended function through moving parts or achange in configuration, are not in the scope of this program.Structural inspections are conducted under the Seabrook Station lifting systems manual.Periodic inspections are conducted at the frequencies, and include the applicable items,delineated in ANSI 830.2, "Overhead and Gantry Cranes," ANSI B30.1 1, "Monorails andUnder hung Cranes," ANSI 830.16, " Overhead Hoists (Under-hung)," and ANSI 830.17,"Overhead and Gantry Cranes (Top Running Bridge, Single Girder, Under-hung Hoist)"for a periodic inspection and in accordance with the manufacturer's recommendations.Inspections are conducted yearly. All periodic inspections are documented on workorders.The enhancement to the program includes:1. The Seabrook Station lnspection of Overhead Heavy Load and Light Load(Related to Refueling) Handling Systems Program Lifting System Manualwill beenhanced to include monitoring of general corrosion on the crane and trolleystructural components and the effects of wear on the rails in the rail system;2. The Seabrook Station Inspection of Overhead Heavy Load and Light Load(Related to Refueling) Handling Systems Program Lifting Systems Manualwill beenhanced to list additional cranes related to the refueling handling system.Enclosure
58.2.1.16 Fire Water SvstemThe Fire Water System Program is an existing program modified to manage the effects ofaging on fire water system components through detailed inspections. Specifically, theprogram manages the following aging effects: loss of material due to general, crevice,pitting, galvanic, and microbiologically influenced corrosion; fouling; and reduction of heattransfer due to fouling of the Fire Water System components.The Seabrook Station Fire Water System Program manages aging of the followingsystem components: sprinklers, nozzles, fittings, filters, valves, hydrants, hose stations,flow gages and flow elements, pumps, standpipes, aboveground and underground pipingand components, water storage tanks, and heat exchangers.The Seabrook Station Fire Protection Manual incorporates activities, such as inspections,flushing, and testing, that serve to prevent or manage aging of the fire water system.Specific examples include: inspections of fire hydrants, fire hydrant hose hydrostatictests, gasket inspections, and fire hydrant flow tests.The Seabrook Station procedures are being enhanced to require the following:inspection sampling or replacing of sprinklers after 50 years of service, flow testing of thefire water system in accordance with National Fire Protection Association (NFPA) 25guidelines, and periodic visual or volumetric inspection of the internal surface of the fireprotection system.The team interviewed the system engineer to understand historical and current conditionsof the system. The team reviewed the current program and existingmaintenance/surveillance procedures to verify the adequacy of the program for detectingand managing aging effects. The team reviewed condition reports to verify that all knownaging effects will be managed by the new program. The team conducted a walkdown ofaccessible portions of system including the electrical penetration area, cable spreadingroom, water storage tanks, and fire pumps to assess the material condition of theaccessible fire water system piping.Based on questions from the team, the applicant modified the application to specify thatthe flow testing will be done in accordance with the 2002 version of NFPA 25. (LicenseRenewal Application change letter SBK-L-1 1063). Also, based on questions from theteam, the applicant modified the application to correct the types of fire water buriedpiping. The original application stated that the fire water piping was polyvinylchloride andcarbon steel. The correct materials were determined to be fiberglass and carbon steel(License Renewal Application change letter SBK-L-1 1054).lEnclosure
6B.2.1.17 Aboveqround Steel TanksThe Aboveground SteelTanks aging management program is a new program used tomanage the aging effects of the external surfaces of five aboveground steel tanks withinthe scope of license renewal. The five tanks within scope are:r Auxiliary Boiler Fuel Oil Storage Tank 1-AB-TK-29o Fire Protection Fuel OilTank 1-FP-TK-3S-Ar Fire Protection Fuel OilTank 1-FP-TK-3S-Bo Fire Protection Water Storage Tank 1-FP-TK-36-Ao Fire Protection Water Storage Tank 1-FP-TK-36-BThe Auxiliary Boiler Fuel Oil Storage Tank 1'AB-TK-?9 has been abandoned. lt isincluded in the application as part of the planning to renovate the tank and return it toservice. All the tanks have protective coatings. The Fire Protection Water Storage Tanksare placed on a concrete pad, leveled using oiled sand, and the edges caulked.The inspector walked-down each of the above tanks. The path chosen by NextEra tomonitor this area was tank level monitoring. For example, blistered paint, with rust andrust stains was noted on Fire Protection Storage Tanks. The tank bottom to concrete padintersection was caulked; however, there was evidence of cracking and peeling of thecaulk. Moisture was present at this intersection and it was not possible to tell if the waterwas from the tank or local inclement weather conditions. The inspector verified theblistered paint with rust, and rust staining was noted in the corrective action program.The inspector also determined, as evidenced by the documented results, that dailyoperator surveillance included the water level of the Fire Protection Storage Tanks. lf themoisture at the bottom of the tank represented a leak, it would be reflected in anunanticipated change in level.The Aboveground Steel Tanks program is credited with managing loss of material on thetank external surfaces including the exterior bottom surface of tanks that is not accessiblefor direct visual inspection. The outer surfaces of the tanks, up to the surface in contactwith the concrete foundation, are managed by visual inspection. Ultrasonic thicknessgauging will be used to monitor loss of material on the inaccessible tank bottom externalsurfaces.8.2.1.20 One-Time lnspectionThe One-Time Inspection Program is a new, one-time program for Seabrook Station thatwill be implemented prior to the period of extended operation. The program will verify theeffectiveness of other aging management programs, including Water Chemistry, Fuel OilChemistry, and Lubricating OilAnalysis Programs, by reviewing various aging effects forimpact. Where corrosion resistant materials and/or non-corrosive environments exist, theOne-Time Inspection Program is intended to verify that an aging management program isnot needed during the period of extended operation by confirming that aging effects arenot occurring or are occurring in a manner that does not affect the safety function ofsystems, structures, and components. Non-destructive examinations will be performedEnclosure
7by qualified personnel using procedures and processes consistent with the approvedplant procedures and appropriate industry standards.The team reviewed application section 8.2.1.20, results of the NRC aging managementprogram audit, and applicant responses to requests for additional information (RAls).The team reviewed the aging management program basis document and draftimplementing guidance, discussed the planned activities with the responsible staff,including sampling plan, and reviewed a sample of corrective action program documentsfor applicable components.8.2.1.21 Selective Leachino of MaterialsThe Selective Leaching of Materials Program is a new, onetime program for SeabrookStation that will be implemented prior to the period of extended operation. The programis credited with managing the aging of components made of gray cast iron, copper alloyswith greater than 15olo zinc, and aluminum bronze with greater than 8% aluminum,exposed to raw water, treated water, and soil environments, which may lead to theselective leaching of material constituents, e.9., graphitization and dezincification. Theprogram will include a one-time visual inspection and hardness measurement test ofselected components that may be susceptible to selective leaching to determine whetherloss of material due to selective leaching is occurring, and whether the leaching processwill affect the ability of the components to perform their intended function during theperiod of extended operation. ln 1998 Seabrook operating experience identified selectiveleaching on aluminum bronze components in sea water. As such, Seabrook will includeperiodic inspections for selective leaching of aluminum bronze as part of this agingmanagement program.The team reviewed application section 8.2.1.21, results of the NRC aging managementprogram audit, and applicant responses to requests for additional information (RAls).The team reviewed the aging management program basis document and draftimplementing guidance, discussed the planned activities with the responsible staff,including sampling plan, and reviewed a sample of corrective action program documentsfor applicable components and for corrective actions to the selective leaching ofaluminum bronze.B.2.1.22 Buried Pipinq and Tanks InspectionThe Seabrook Station Buried Piping and Tanks Inspection Program is a new programthat includes coating, cathodic protection, and backfill quality as preventive measures tomitigate corrosion. Periodic inspections manage the aging effects of corrosion on buriedpiping in the scope of license renewal. Buried steel and stainless piping has an externalprotective coating consisting of coal-tar primer, coal-tar enamel, asbestos felt or fibrousglass mat, and a wrapping of kraft paper or coat of whitewash. Some hot-applied tapecoating was also used. Coatings were fabricated and applied in accordance with therequirements of American Water Works Association specification C203 and this required"holiday" (flaws in coating) testing.Enclosure
8Backfill was applied in accordance with Seabrook Specification 9763-8-1, "Bedding,Backfilling and Compaction for Miscellaneous Non Safety Related Piping" and 9763-8-5"Bedding, Backfilling and Compaction for Safety Related Systems and Structures".Except for the allowance of backfill at a size of 1/z" the backfill is equal to or better thanthe GALL Revision 2 proposal of ASTM D 448-08 Size 67. As a consequence, NextErais proposing inspection in conformance with an acceptable backfill limit until a discoveryis made of coating damage. For steel with cathodic protection, they propose 1inspection. lf backfill damage is discovered, they will increase this by another 3 samples.For steel without cathodic protection, they propose 4 inspections; and if backfill damageis discovered, they will expand by another 4 inspections.The team reviewed cathodic protection system reports and determined the system was indisrepair since being identified as unreliable in 1993. The system was not restored until2007 when a survey found that only 620/o of the areas surveyed were being mitigated bycathodic protection. During the first quarter of 2009 the cathodic protection system wasfinally categorized as green (or satisfactory condition). The cathodic protection systemwas made a Maintenance Rule (10 CFR 50.65) System during the same quarter.There is an adequate historical basis to conclude that buried piping was adequatelyprotected, and the backfill correctly specified and filled, during construction. There is anabsence of buried piping problems at the site. Because there was an absence of aconsistent cathodic protection for a period of 1993 to 2009, it is appropriate for NextEra toinspect buried piping by excavation to corroborate the historical basis.B.2.1.23 One-Time Inspection of ASME Code Class 1 Small Bore PipinqThe One-Time Inspection of ASME Code Class 1 Small Bore Piping Program is a newprogram that manages the aging effect of cracking in stainless steel small-bore ASMECode Class 1 piping less that 4 inches nominal pipe size, including pipe, fittings, andbranch connections. Seabrook has not experienced a small bore piping failure due tostress corrosion or thermal and mechanical loading. The small bore piping selected forinsonification is based on EPRI Report 1011955, "Management of Thermal Fatigue inNormally Stagnant Non-lsolable Reactor Coolant System Branch Lines (MRP-146)",issued June 2005 and the supplementalguidance issued in EPRI Report 1018330,"Management of Thermal Fatigue in Normally Stagnant Non-isolable Reactor CoolantSystem Branch Lines - Supplemental Guidance (MRP-1465) issued December of 2008.Using these criteria the applicant has identified 448 welds, of which 157 are socket welds(including 58 in-core instrument guide tube welds) and 291 butt welds. In this populationthere are 6 small bore stagnant segments susceptible to thermalfatigue. These are inthe two charging lines and four high head safety injection lines. These locations aremonitored.Twenty-Nine (29) welds (4 socket and 25 butt welds) have been identified in the 448candidates as vulnerable to cracking. These will be tested using ultrasonic inspection notsooner than 10 years before the extended period of operation.Enclosure
9B.2.1.25 Inspection of lnternal Surfaces in Miscellaneous Pipinq and DuctinoComponentsThe Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components(lnternal Surfaces) Program is a new program that will inspect the internals of piping,piping components, ducting, and other components of various materials to manage theaging effects of cracking, loss of material, reduction of heat transfer, and hardening ofelastomers. The inspections of opportunity will occur during maintenance andsurveillance activities when systems are opened.The team reviewed application section 8.2.1.25, draft NRC aging management programaudit, and applicant responses to requests for additional information (RAls). The teamreviewed the aging management program basis document, operating experience reviewdocuments, draft implementing guidance, and relevant condition reports. The teaminterviewed applicable plant personnel.B.2.1.26 Lubricatinq Oil AnalvsisThe Lubricating OilAnalysis Program is an existing program, which maintains oil systemsfree of contaminants (primarily water and particulates), thereby preserving anenvironment that is not conducive to loss of material, cracking, or fouling. The applicantperforms sampling, analysis, and trending of results on numerous systems to provide anearly indication of adverse equipment condition in the lubricating oil environment. Theapplicant samples the lubricating oil for most of the affected equipment on frequenciesrecommended by the vendor.The team reviewed application section 8.2.1.26, draft NRC aging management programaudit, and applicant responses to requests for additional information (RAls). The teamreviewed the aging management program basis document, operating experience reviewdocuments, existing procedures, relevant condition reports, and system health reports.The team interviewed plant personnel and sampled oil measurement results and trendingwithin the applicant's database. Further, the team performed walk downs of thelubricating oil components of B emergency diesel generator.The team identified an issue regarding the existing lubricating oil practice on testing forwater content. Specifically, the applicant tests for water content on lubricating oil forpumps and motors when these components are water-cooled and have the potential forwater contamination. Nonetheless, the team identified that the lubricating oil andhydraulic fluid samples of charging pump 1-CS-P-128 were not being tested for watercontent despite the pump being water-cooled. The applicant issued Action Request01632769 to correct the testing for water content on this pump, to confirm test packagesfor other components are correct, and to review the testing for water content of all pumpsand motors as part of the enhancement to the program to provide a program attachmentwith the required equipment and the specified sample analyses and frequency.Enclosure
10B.2.1.27 ASME Section Xl. Subsection IWEThe ASME Section Xl, Subsection IWE aging management program is an existingprogram, credited in the LRA, which provides for inspecting the reactor building liner plateand related components for loss of material, loss of pressure retaining bolting preload,cracking due to cyclic loading, loss of sealing, and leakage through seals, gaskets andmoisture barriers in accordance with ASME Section Xl. Areas of the reactor buildingadjacent to the moisture barrier and the moisture barrier are subject to augmentedexamination.The team reviewed applicable procedures, the latest lnservice Inspection program resultsand interviewed the Inservice lnspection program manager. The team reviewed asample of recent corrective action reports from Section IWE examinations.The team concluded that the Inservice Inspection program was in place, had beenimplemented, was an on-going program subject to NRC review, and included theelements identified in the license renewalapplication.8.2.1.28 ASME Section Xl. Subsection IWLThe Seabrook Station ASME Section Xl, Subsection IWL Program is an existing programthat manages the aging effects of cracking, loss of bond, loss of material (spalling,scaling) due to corrosion of embedded steel, expansion and cracking due to reaction withaggregates, increase in porosity and permeability, cracking, loss of material (spalling,scaling) due to aggressive chemical attack, and increase in porosity and permeability,loss of strength due to leaching of calcium hydroxide.The team reviewed the program basis documents, program description, baselineinspection results, subsequent inspection results for trending, and implementingprocedures to determine the scope and technical adequacy of the Program. Also, theteam reviewed ARs to assess the adequacy of evaluations of findings, and resolution ofconcerns, if any, identified in these inspections.The team observed that the program complies with the requirements of ASME Section Xl,Sub-Section lWL, "Requirements for Class CC Concrete Components of Light-WaterCooled Power Plants". The components examination contained in 10 CFR 50.55a inaccordance with ASME Boiler and Pressure Vessel Code, Section Xl, Subsection IWLmanaged by the program include steel reinforced concrete for the Seabrook Stationcontainment building and complies with the requirement for examination contained in 10CFR 50.55a in accordance with ASME Boiler and Pressure Vessel Code, Section Xl,Subsection lWL.The primary inspection method used at Seabrook Station is W-1C visual examination,W-3C visualexamination, and alternative examination methods (in accordance withIWA-2240). The Seabrook Station ASME Section Xl, Subsection IWL Program providesacceptance criteria and corrective actions for each exam type. The team noted, for thisprogram and the structures monitoring program, a technically acceptable trending systemwas not implemented to establish the status of observed cracks (stable or active), andIEnclosure
11qualification and certification of inspectors/examiners was not explicitly established anddocumented to assure assignment of qualified individuals for inspection. The inspectionpersonnel selection is left to the supervisor of the group. Also, there was a lack of clearquantitative acceptance/evaluation criteria established by the procedure to assureconsistency in observation, evaluation, and assessment of inspection results by differentinspectors and technical personnellengineers and at different times. This program will befurther enhanced with revised implementing procedures to include definition of"Responsible Enginee/'(letter SBK-L-10204, RAl 8.2.1.28-3, Commitment No. 31) andtrending information and acceptance criteria (same letter, RAI 8.2.1 .28-1).Concrete degradation due to alkai-silica reaction is an aging effect that wasrecentlydiscovered for Seabrook Station. In addition to the control building, it had beennoted in other buildings such as Emergency Diesel Generator Building, and the ResidualHeat Removal Vault (see additional details in the section b of this report). The Teamreviewed applicant photographs of pattern cracking on the primary containment wall inthe annulus region. The annulus region appears to have had approximately six feet ofwater for an extended period of time due to groundwater infiltration. NextEra plans tokeep the area drained (Letter SBK-L-11063 commitnment No. 52) and to review, analyze,and assess the effect of this condition in order to determine the cause on the primarycontainment (AR 01641413, Crazed Crack Pattern On Containment In Annulus Area).8.2.1.31 Structures Monitorinq ProoramThe Structures Monitoring Program at Seabrook Station is an existing program that is tobe further enhanced to be consistent with guidance set forth in 10 CFR 50.65,"Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants",NUMARC 93-01, "lndustry Guidelines for Monitoring the Effectiveness of Maintenance atNuclear Power Plants", and Regulatory Guide 1.160, Rev. 2, "Monitoring theEffectiveness of Maintenance at Nuclear Power Plants". This program is described inAppendix B, Section 2.39 tor the license renewal application. The applicant uses thestructural monitoring program to monitor the condition of structures and structuralcomponents within scope of the Maintenance Rule, thereby providing reasonableassurance that there is no loss of intended function of structure or structural component.As noted in the application, the program will be enhanced to include: additional structuresand structural components identified in the license renewalaging management review,add aging effects, additional locations, inspection frequency, and ultrasonic testrequirements and enhancements for procedures to include inspection opportunities whenplanning excavation work that would expose inaccessible concrete. Enhancements tothe Structural Monitoring Program will be implemented prior to the period of extendedoperation.Aging effects or material degradation in concrete identified within the scope of theStructures Monitoring Program such as loss of material, cracking, change in materialproperties, and loss of form are detected by visual inspection of external surfaces prior tothe loss of the structure's or component's intended function.The team reviewed the Aging Management Program description for the StructuralMonitoring Program, the Program Evaluation Document for the Structural MonitoringProgram, engineering documents, inspection reports, condition reports, corrective actionEnclosure
12documents, work request documents, site procedures, and related references used tomanage the aging effects on the structures. During the inspection the team conducted ageneral walkthrough inspection of the site, including the turbine building, reactorcontainment building, diesel generator building, control room, the intake structure, andother applicable structures, systems, and components related to the Structural MonitoringProgram. The team held discussions with applicant's supervisory and technicalpersonnel to verify that areas where signs of degradation, such as spalling, cracking,leakage through concrete walls, corrosion of steel members, deterioration of structuralmaterials and other aging effects, had been identified and documented. Also, the teamverified that the applicant maintains appropriate (photographic and/or written)documentation of these inspections to facilitate effective monitoring and trending ofstructural deficiencies and degradations.Through the review of documents, walkthrough inspections, and discussions withengineering and plant personnel, the inspector identified some weaknesses in thestructural aging management program. Similar to the IWL program, the inspectorobserved the need for clarification on acceptance criteria and the responsible engineerperforming inspections. The applicant agreed to the needed changes as noted in the IWLprogram 8.2.1.27 (previous section).As noted in the IWL program, concrete degradation due to alkai-silica reaction is an agingeffect that was recently discovered for Seabrook Station (see additional details in thesection b of this report).8.2.1 .32 Electrical Cables and Connections Not Subiect to 10 CFR 50.49 EQRequirementsThe Electrical Cables and Connections Not Subject To 10 CFR 50.49 EnvironmentalQualification Requirements Program is a new program that will manage the aging effectsof embrittlement, cracking, discoloration or surface contamination leading to reducedinsulation resistance or electrical failure of accessible cables and connections due toexposure to an adverse localized environment caused by heat, radiation or moisture inthe presence of oxygen. This program applies to accessible cables and connectionsinstalled in in-scope structures.This program will visually inspect accessible electrical cables and connections installed inadverse localized environments at least once every 10 years. The first inspection forlicense renewal is to be completed before the period of extended operation. An adverselocalized environment is defined as a condition in a limited plant area that is significantlymore severe than the specified service environment (i.e. temperature, radiation, ormoisture) for the cable or connections.The team conducted walkdowns to observe cable and connector conditions in potentialadverse localized environments. The team reviewed condition reports and interviewedplant personnelto assess historical and current conditions. The team reviewed the draftprogram documents to verify the program will be able to manage aging effects.Enclosure
138.2.1.34 Inaccessible Power Cables Not Subiect To 10 CFR 50.49 EQ RequirementsThe Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental QualificationRequirements Program is a new program that will manage the aging effects of localizeddamage and breakdown of insulation leading to electricalfailure of inaccessible powercables (400V and higher) due to adverse localized environments caused by exposure tosignificant moisture. Seabrook Station defines an adverse localized environment forpower cables as exposure to moisture for more than a few days.The Seabrook Station program includes periodic inspections of manholes containing in-scope cables. The inspection focuses on water collection in cable manholes, and drainingwater, as needed. The frequency of manhole inspections for accumulated water andsubsequent pumping will be based on plant specific operating experience, The maximumtime between inspections will be no more than one year.ln addition to periodic manhole inspections, in-scope cables are tested to provide anindication of the condition of the conductor insulation. The specific type of test performedwill be determined prior to the initial test, and is a proven test for detecting deterioration ofthe insulation system due to wetting, such as power factor, partial discharge, orpolarization index or other testing that is state-of-the-art at the time the test is performed.Cable testing will be performed prior to entering the period of extended operation and atleast every six years thereafter.Overall actions are to test cables and keep them dry. Seabrook has had, and continuesto get, some water in their manholes. NextEra is taking corrective actions by increasingthe inspection frequency and pumping frequently to prevent submergence of safety-related cables. They are committing to having initial inspections done and adjustinspection/pumping frequencies based on experience.The team interviewed the responsible system engineer to understand the proposedprogram and power cable operating experience at Seabrook. The team reviewed datafrom previous manhole inspections to verify the established inspection frequencies arecommensurate with operating experience. The team observed the inspection of a below-ground manhole at Seabrook to assess the process for inspections and the materialcondition of the manhole. The team reviewed system health reports and conditionreports for historical operating experience and program guidance for cable conditionmonitoring to assess the adequacy of the proposed program to manage aging effects.B.2.1.35 Metal Enclosed BusThe Metal Enclosed Bus Program is a new program that will manage the following agingeffects of in-scope metal enclosed buses: loosening of bolted connections due to thermalcycling and ohmic heating; hardening and loss of strength due to elastomer degradation;loss of material due to general corrosion; and embrittlement, cracking, melting, swelling,or discoloration due to overheating or aging degradationThis new program will be implemented prior to entering the period of extended operationand at least once every 10 years thereafter.Enclosure
14The internal portions of the in-scope metal enclosed bus enclosures will be visuallyinspected for aging degradation of insulating material and for cracks, corrosion, foreigndebris, excessive dust buildup, and evidence of moisture intrusion. The bus insulationwill be visually inspected for signs of embrittlement, cracking, melting, swelling, ordiscoloration, which may indicate overheating or aging degradation. The internal bussupports will be visually inspected for structural integrity and signs of cracks. Theaccessible bus sections will be inspected for loose connections using thermography fromoutside the metal enclosed bus through the viewport, while the bus is energized.The team reviewed previous work orders for inspection and cleaning activities for metalenclosed buses. The team interviewed the associated system engineer and reviewedcondition reports to assess the historical and current condition of the metal enclosedbuses. The team conducted a walkdown of accessible portions of the metal enclosedbuses to evaluate the exterior condition of the buses and the operating environment.8.2.2.1 34 5 kV SFG BusThe Seabrook Station 345kV SF6 Bus Program is a new plant-specific program that willmanage the following aging effects on the 345kV SF6 Bus: loss of pressure boundarydue to elastomer degradation; loss of material due to pitting; crevice and galvaniccorrosion; and loss of function due to unacceptable air, moisture or sulfur dioxide (SOz)levels.Sulfur Hexafluoride (SF6) is an inert gas used to insulate bus conductors. The programwill inspect for corrosion on the exterior of the bus duct housing, test for leaks atelastomers, and periodically test gas samples to determine air, moisture and SOz levels.Inspections, leak testing, and gas sampling will be done prior to entering the period ofextended operation and at least once every six months thereafter.The team reviewed previous work orders for maintenance activities associated withinspections of the SF6 buses and SFo gas monitoring. The team interviewed theassociated system engineer and reviewed condition reports to assess the historical andcurrent condition of the SFo buses. The team reviewed system health reports to verifythat any aging effects are being adequately managed. The team conducted a walkdownof the SF6 buses to evaluate the exterior condition of the buses and the operatingenvironment.B.2.2.2 Boral MonitorinqThe Boral Monitoring Program is an existing program used to monitor the condition of thematerial used in spent fuel pools for reactivity control. Boral is the brand name for asheet of uniformly distributed boron carbide in an alloy 1 100 aluminum matrix with a thinaluminum clad on both sides. The predecessor to Boral is Boraflex, a similar materialsusceptible to radiolytic degradation. Boraflex is used in the first six sets of racks atSeabrook. The Boraflex utilized in the initial six racks is not credited in the criticalityanalyses and is not credited for license renewal.Enclosure
15The aging affect requiring management is a reduction in neutron absorbing capacity, achange in dimensions, and a loss of material due to the affects of the spent fuel poolenvironment. Boral exposed to treated borated water is the subject of Draft LR-ISG-2009-01, "Staff Guidance Regarding Plant Specific Aging Management Revieft andAging Management Program for Neutron-Absorbing Material in Spent Fuel Pools"The team reviewed the program documents, reviewed various corrective actions, andinterviewed the responsible engineers.B.2.2.3 Nickel-Allov Nozzles and PenetrationsThe Nickel-Alloy Nozzles and Penetrations Program is an existing program that managescracking, due to primary water stress corrosion, of the nickel based alloy pressureboundary and structural components exposed to the reactor coolant. This includesPressurizer Nozzles, Steam Generator Channel Head Drain Tube and Welds, ReactorVessel Core Support Pan/Lug, and Clevis Inserts, Reactor Vessel Hot and Cold LegNozzles, and the Reactor Vessel Bottom Mounted lnstrumentation Penetrations. Theprogram has been in existence, in various forms, since 2004 when Seabrook respondedto NRC Bulletin 2004-01 "lnspection of Alloy 8211821600 Materials Used in theFabrication of Pressurizer Penetrations and Steam Space Piping Connections atPressurized Water Reactors". The management of this aging affect has been refinedsince the phenomena was first described and has culminated in the Electric PowerResearch lnstitute sponsored program MRP-139 "Material Reliability Program: PrimarySystem Piping Butt Weld lnspection and Evaluation Guideline".Seabrook's draft "Reactor Coolant System Materials Degradation Management Program"is structured around the primary goal of mitigating material degradation of the reactorcoolant system pressure boundary and reactor vessel internals. The program is intendedto manage the "Steam Generator Program", Thermal Fatigue Management Program","Alloy 600 Program", "Boric Acid Program", "Reactor Vessel lnternals Program", and the"ASME Section Xl Program (NDE, lSl, Repair/Replacement)". The management programincludes an appendix titled "Westinghouse Proprietary Information", which identifiespotential Alloy 600/821182locations in the primary pressure boundary components of theWestinghouse designed Nuclear Steam Supply System.Svstem ReviewIn distinction to the above noted program review, a system review was chosen by theteam as a different approach to ensure comprehensive coverage of aging effects. TheResidual Heat Removal System was chosen since the most likely initiating event, atSeabrook, is a station black out and a dominate system for station black out response isthe Residual Heat Removal System. The approach is to walk down the system in theplant and question how aging effects are covered and verify that coverage based on areview of the application, program descriptions, and if available implementing procedures.Materials identified for this system are Cast Austenitic Stainless Steel, Glass, StainlessSteel, and Steel in the external environments of indoor air that may included borated andEnclosure
16non-borated water leakage and Closed Cycle Cooling Water. The internalenvironmentsare various treated and untreated water, lubricating oil, and reactor coolant.This results in the possible or experienced aging affects of cracking, (cyclic, stresscorrosion, thermal, loaded, and fatigue) and corrosion (boric acid, crevice, galvanic,general, and pitting), loss of preload, and fouling.The applicant, in turn, proposes the following aging management programs:ASME Section Xl Subsections lWB, lWC, and IWD ProgramBolting Integrity ProgramBoric Acid ProgramClosed-Cycle Cooling Water System ProgramExternal Surfaces Monitoring ProgramLubricating Oil Analysis ProgramOne'Time Inspection of ASME Code Class Small Bore PipingOne-Time Inspection ProgramWater Chemistry ProgramThe ASME Section Xl Subsections lWB, lWC, and IWD program, the Boric Acid Programare reviewed at every outage under the NRC's Reactor Oversight Program usinginspection procedure 1P71111.08P "lSl Inspection". The Water Chemistry Program ispart of the same procedure by way of the Steam Generator inspection portion. TheBolting Integrity Program, One-Time Inspection of Code Class Small Bore Piping, andOne-Time lnspection are covered elsewhere in this report.Of interest was a note in the System Walk-down Report, in 2008, recording the presenceof water intrusion associated with "several supports in the vault stairuvell" and theobservation the "conditions are slowly becoming worse as calcium accumulates." WO0844358 was initiated to verify the bolting integrity. The work order incorrectly comparedthe testing of anchors submerged in raw water in a manhole with the anchors supportingthe RHR piping inserted into a calcium carbonate degraded wall and concluded, basedon the submerged bolting, that the bolting in the RHR anchors were acceptable (AR01633206). This comparison did not take into account the additional concern of arecently discovered alkaline silica degradation associated with the calcium carbonatedegraded wall and the issue of anchor bolting integrity was not revisited subsequent tothe discovery of alkali silica degradation. WO 0844358 was translated, during a databasechange, into Condition Report 08-15902 and closed on the basis of the comparison (twodifferent material environmental conditions) even though the condition report contained aproposal to randomly sample the bolts and perform a calibrated torque test. Theimplications of the NRC BulletinT9-02 anchor bolt integrity program were neverconsidered during the evolution. lnitially, these erroneous comparisons, and incompleteanalysis, indicate a weakness in the NextEra's program for identifying and tracking therecently discovered aging effects at the site. The revised analysis resulted in satisfactoryconditions and the learning needed in dealing with aging effects to support licenserenewal (AR 01633206).Enclosure
b.17The inspector walked-down the RHR system from the outlet of RHR Pump P-8A, atelevation 54"-4", to the inlet of RHR Heat Exchanger E-gA, at elevation -31"-0", pausingat each support to carefully inspect the visual appearance of the bare piping revealed bythe gaps in insulation. The inspector did not identify any evidence of aging that was notalready considered by the applicant and adequately covered by an existing of proposedprogram.Observations and FindinqsAlkali-Silica Reaction Aqinq Effect at Seabrook StationTo assess the material condition of concrete structures in the plant; and to acquire, verify,and validate the design basis of structural design, the applicant personnel performedcivil/structuralwalk-down inspections. The Residual Heat Removal Equipment Vaults, Aand B Electrical Tunnels, Radiological Controlled Area Walkway, and Service Waterpump house was included in the walk-down inspection and assessment. Theobservations and findings were documented in the License Renewal Project issuetracking report number 15. The walk-down inspections discovered the following plantmaterial conditions; (a) large amount of groundwater infiltration, (b) large amount ofcalcium carbonate deposits, (c) corroded steel supports, base plates and piping,(d) corroded anchor bolts, (e) pooling of water and (f) cracking and spalling of concrete.The inspection further noted that the below grade, exterior walls in the Control Building BElectrical Tunnel at elevation (-) 20'- 00" have random cracking and for several years havebeen saturated by groundwater infiltration. The severity of the cracking and groundwaterinfiltration varies from location to location. The groundwater infiltration has produced large,tightly adherent deposits of calcium oxide/carbonate at certain locations on the walls andpooling of groundwater on the floor slab sometimes to a depth of 2-inches. Thegroundwater has also produced smaller, loose deposits of calcium salts at most other cracklocations.The observations and findings from the walk-down inspections were reviewed byapplicant's Design Engineering Organization and it was determined that the concretewalls in the B-Electrical Tunnel exhibited the most extensive distressed condition asdetermined by the applicant and required further investigation. Specifically, the belowgrade exterior walls in the Control Building B Electrical Tunnel at elevation (-) 20' - 00" wereselected due to the presence of fine, random cracking and, because, for over 10 to 15years had remained in saturated condition by groundwater infiltration. The severity of thecracking and groundwater infiltration varied from location to location. The groundwaterinfiltration had produced large, tightly adherent deposits of calcium oxide at certainlocations on the walls and pooling of groundwater on the floor slab sometimes to a depth of2-inches. The groundwater has also produced smaller, loose deposits of calcium oxide atmost other crack locations.To assess the integrity of cracked concrete and prolonged groundwater saturation, theapplicant contracted with vendors to perform Penetration Resistance Testing (also referredto as Windsor Probe Test), and also to obtain concrete core specimens at designatedlocations in four below grade, exterior walls of the B Electrical Tunnel. The concrete coreEnclosure
18specimens were subjected to compressive testing by the vendor and selected sections ofthe core specimens were provided to another vendor for Petrographic examinations.The results Penetration Resistance Tests (PRT) for the control building indicated anaverage concrete compressive strength of 5340 psi and the concrete core testingindicated an average compressive strength of 4790 psi. PRT performed in 1979indicated an average concrete compressive strength of 6750 psi and the concrete testcylinders that were cast during the placement of the walls in February 1979 indicated anaverage 28-day compressive strength of 6120 psi. At each of the six (6) locations, three(3) individual replicate Penetration Resistance Tests as recommended per ACI 228.1R,Tables 5.2 and 5.5 has been performed for a total of eighteen (18) Penetration ResistanceTests. Each of the eighteen (18) PRTs required three (3) firmly embedded probes asrecommended in ASTM C 803-03, paragraph 8.1.2for a total of fifty-four (54) probes. ThePRTs shall be performed per ASTM C 803-03 standard, utilizing Windsor Probe TestSystem per foreign print no. 100561.At each of six (6) locations, core drilled and removed two (2), 4-inch nominaldiameterconcrete core specimens as recommended in ACI 228.1R, paragraph 4.3.2. A totaloftwelve (12) concrete core specimens will be obtained as recommended in ACI 228.1Rparagraph 4.3.2to develop an adequate strength relationship between the PRTs and thein-situ compressive strength of the concrete. The concrete core specimens has beenobtained per the method specified in ASTM C 42-04 and compression tested in the ME&Tlaboratory per ASTM C 39-09. The length of the concrete core specimens "as removed"were12 to 16-inches maximum. This provided adequate specimen lengths for compressiontesting and Petrographic examinations. All of the walls in the B Electrical Tunnel includedin this study were 2-foot in thicKness per drawing 101345, thus the concrete core drilling didnot penetrate through the walls or contacted the two layers of reinforcement on the outer-face of the walls.A comparison of the 2010 concrete compression test results to the 1979 concretecompression test results indicated a 21.7 percent reduction in the compressive strengthof the concrete. The reduction in compressive strength is most likely due to alkali-silicareaction in the concrete which was detected in Petrographic examinations of four of theconcrete core samples removed from the CB walls. lt was reported that the four concretecore samples had moderate to severe Alkali-Silica Reaction in the concrete. Alkali-SilicaReaction is a reaction that occurs over time in concrete between the alkaline cementpaste and reactive non-crystalline silica which is found in many common coarseaggregates. The reaction produces a gel substance which expands and causes micro-cracking or fissures in and surrounding the coarse aggregates. The micro-crackingtypically progresses and extends into the cement paste thus compromising the qualityand integrity of the concrete. The presence of water, irrespective of water chemistry (i.e.,aggressive or non-aggressive), is required for Alkali-Silica Reaction to develop and tocontinue to propagate in the hardened concrete. Without the presence of water, Alkali-Silica Reaction will not develop or continue to propagate in hardened concrete. Alkali-Silica Reaction often results in a reduction in both strength and elasticity of the concrete;both of which were noted in the sample concrete cores analyzed for Seabrook.Enclosure
19The reduction in compressive strength raises questions regarding the effect on modulus ofelasticity, and flexural and shear capacity of concrete structural members. ln addition themodulus of elasticity affects the dynamic response of Structures. The applicant isconsidering the structure dynamic response in their analyses.In accordance with Inspection Procedure 71002 and Inspection Manual Chapter 2516, akey assumption of license renewal is that the current licensing bases is to be maintained.The above discussion indicated that this may not be true if operability of the safety relatedstructures cannot be maintained. The NRC inspection report 0500044312011002, issuedMay 12,2011, addresses current licensing bases issues along with an extent of conditionreview planned by the applicant.With respect to the aging management review for this aging effect at the station, theapplicant provided a summary of their plans in a response for additional informationassociated with the Division of License Renewal review in a letter datedApril 14, 2011 (letter SBK-L-11063).Overall FindinosThe team concluded screening and scoping of non-safety related systems, structures,and components, was implemented as required in 10 CFR 54.4(a)(2), and the agingmanagement portion of the license renewal activities were conducted as described in theLicense Renewal Application. The inspection concluded the documentation supportingthe application was in an auditable and retrievable form. Except for the alkali-silicareaction issue, the inspection results support a conclusion of reasonable assurance withrespect to managing the effects of aging in the systems, structures, and componentsidentified in the application.Enclosure
A-1ATTACHMENTSUPPLEMENTAL INFORMATIONKEY POINTS OF CONTACTApplicant PersonnelE. Metcalf Plant ManagerM. Collins Design Engineering ManagerM. O'Keefe Seabrook Station Licensing ManagerR. Cliche License Renewal Project ManagerP. Tutinas License Renewal Project Electrical LeadA. Kodal License Renewal Project Mechanical LeadK. Chew License Renewal Project CivilStructural LeadLIST OF DOCUMENTS REVIEWEDGeneral License Renewal DocumentsNRC lnspection Procedure 71002; License Renewal InspectionNRC AMP Audit Report (results)SBK-L-10192, Seabrook Station, Response to RAls, Set ?, X,2Q10SBK-L-10204, Seabrook Station, Response to RAls, Set ?, December 17 ,2Q10SBK-L-11002, Seabrook Station, Response to RAls, Set 4, January 13,2011SBK-L-11003, Seabrook Station, Response to RAls, Set 5, January 13,2011SBK-L-11015, Seabrook Station, Response to RAls, Set ?, X,2011SBK-L-1 1027, Seabrook Station, Response to RAls, Set 9, X,2011Updated Final Safety Analysis Report, Section 3.7(8).3.13License Renewal Basis DocumentsLRAM-ELEC, Aging Management Review Report: Electrical Components and Commodities,Rev 1LRAP-EI, Aging Management Program Basis Document: Electrical Cables and Connections NotSubject to 10 CFR 50.49 Environmental Qualification Requirements, Rev 2 and Rev 3LRAP-E3, Aging Management Program Basis Document: Inaccessible Power Cables NotSubject to 10 CFR 50.49 Environmental Qualification Requirements Program, Rev 2LRAP-E3, Aging Management Program Basis Document: Metal Enclosed Bus, Rev 1LRAP-M027, Aging Management Program Basis Document: Fire Water System, Rev 1LRAP-M032, Aging Management Program Basis Document: One-Time lnspection, Revision 1LRAP-M033, Aging Management Program Basis Document: Selective Leaching of Materials,Revision 1LRAP-M033, Aging Management Program Basis Document: Selective Leaching of Materials,Revision 2 (Draft)IAttachment
A-2LRAP-M038, Aging Management Program Basis Document: lnspection of lnternalSurfaces inMiscellaneous Piping and Ducting Components, Revision 1LRAP-M039, Aging Management Program Basis Document: Lubricating OilAnalysis, Revision 1LRAP-SF6, Aging Management Program Basis Document: 345kV SF6 Bus, Rev 1LRSP-ELEC, Scoping and Screening Report: Electrical Systems, Components, andCommodities, Rev 2LRTR-NSAS, Technical Report - Non-Safety Affecting Safety, Revision 3LRTR-NSAS, Technical Report - Non-Safety Affecting Safety, Revision 4lmplementino ProceduresCP 3.3, Closed Cooling Water Systems, Chemistry Control Program, Rev 20ER-AA-106, Cable Condition Monitoring Program, Rev 1ES1807.020, Machinery OilAnalysis, Revision 0FP 3.1, Fire Protection Maintenance and Surveillance Testing, Rev 3LN0560.10, SFO Dewpoint Check, Rev 21N0560.11, SFO SO2 and Purity Sample, Rev 7ON0443.54, Non-safety Related Deluge and Sprinkler Systems 18 Month lnspection, Rev 4,Change 8AN1242.01, Loss of lnstrumentAir, Revision 12030443.66, Safety Related Spray and Sprinkler Systems 18 Month Flow and System AlarmsTest, Rev 4, Change 9OX0443.04, Fire Protection System Annual Flush, Rev 6 Change IOX0443.12, Fire Protection Dry Pipe Spray and Sprinkler Systems 18 Month Inspection, Rev 6,Change 4OX0443.19, Yard Hydrant Hose House Monthly Inspection, Rev 6 Change 4OX0443.20, Yard Hydrant Semi-Annual lnspection and Functional Test, Rev 6, Change 6OX0443.21, Yard Fire Hydrant Hose Houses Annual Hose Replacement and Gasket lnspection,Rev 6, Change 2PEG'10, System Walkdowns, Rev 18PEG-265, Cable Condition Monitoring, Rev 0SSCP, Chemistry Manual, Rev 64Draft lmplementinq ProceduresLRTR-INT, Technical Report - lnspection of Internal Surfaces, Revision 0 (Draft)LRTR-OTI, Technical Report - One-Time lnspection, Revision 0 (Draft)LRTR-SEL, Technical Report - Selective Leaching of Materials, Revision 0 (Draft)Technical ReportsEE-07-018, Response to GL 2001-01, Rev 0Engineering Evaluationg4-41, Submerged Electrical Cables and Supports, Dated 1l39l95Technical Report "Buried Piping and Tanks lnspection Program" LRTR-BP Revision 0Attachment
A-3Work Orders008088601 819640187223023429502424560301 31 1031 0880031769604016970401699040172804065340414066041758804316570443640044432105199530526073060304247027050716257071625807189940719543072039007271170727135072713607271370727138081 342008270610827184082718508313120831 31 30831583083565698C388999A5575IAttachment
A-4Work Order Package 00611225 01, "Reference Maintenance - Auxilliary Boiler Tank ManwayLeakage"Work Order Package 00616970 01, "The Outside of FP-TK-36A Has Peeling Paint and Rust TK"Work Order Package 00616971 01, "The Outside of FP-TK-368 Has Peeling Paint and Rust TK"Work Order Package 00791046 01, "Diesel Fire Pump Fuel Oil Tank Water Removal"Work Order Package 00791057 01, "Diesel Fire Pump Fuel Oil Tank Water Removal"Action Request 00207755 "Seabrook Station License Renewal lmplementation Actions"Completed Surveillance Tests12 oil sample analysis results from Herguth LabsReference DocumentsMaterials Reliability Program: Primary System Piping Butt Weld Inspection and EvaluationGuidelines (MRP-139) 1010087, August 2005NEI 96-03, Guideline for Monitoring the Condition of Structures at Nuclear Power Plants, 1996ACI 201.1R-92, Guide for Making a Condition Survey of Concrete in Service, American ConcreteInstituteACI 349.3R-96, Evaluation of Existing Nuclear Safety- Related Concrete Structures,American Concrete lnstitute ACI 531-79, Concrete Masonry Structures, Design andConstruction, American Concrete lnstituteHope Creek Update Final Safety Analysis Report, Section 7.2.1.36Materials Reliability Program: Primary System Piping Butt Weld Inspection and EvaluationGuidelines (MRP-139) 1010087, August 2005NEI 09-14, Revision 0; Guidelines For The Management Of Buried Piping lntegrity, 01110EPRI Final Report 1016456, 121Q8; Recommendations for an Effective Program to ControltheDegradation of Buried PipingDrawinosComplete set of submitted license renewal drawings1-AS-2301-2, Auxiliary Steam Piping, Revision 41-AS-5198-02, Auxiliary Steam Piping, Revision 31-DM-D20355, Demineralized Water Distribution Detail, Revision 179763-F-310248, Underground Duct Plan, Rev 139763-F-802807-641.20C, Piping - Combustible Gas lsometric, Revision 09763-F-802807S, Sheets 15, 155, 16; Pipe Support Details, Revision 689763-F-202753-610.60, Service Air lsometric, Revision 09763-M-202751S, Sheets 43, 43S, 74,745,74A; Support Details, Revision 25AAttachment
A-59763-M-212368S, Sheets 15, 155, 16; Support Details, Revision 11B9763-M-212368S, Sheets 17, 175,18, 18A; Support Details, Revision 23A9763-M-2123685, Sheets 19, 195; Support Details, Revision 2089763-M-2123685, Sheets 36, 365, 37; Support Details, Revision 1289763-M-2123685, Sheets 53, 53S, 54 - 57; Support Details, Revision 24A9763-M-8029133, Sheets 49, 49S, 50, 51, 52; Support Details, Revision 11B1-NHY-310002, Unit Electrical Distribution One Line Diagram, Rev 401-NHY-505084, Instrument Air Installation - DualAir Supply, Revision 6PID-1-WLD-820224, Waste Processing Liquid Drains - RCA Walkway Details, Revision 7License Renewal PID Drawing PID-1-RH-1R20663License Renewal PID Drawing PID-1-SI-LR20446License Renewal PID Drawing PID-1-Sl-LR20447License Renewal PID Drawing PID-1-Sl-LR20448License Renewal PID Drawing PID-1-Sl-LR20449License Renewal PID Drawing PID-1-Sl-1R20450License Renewal Pl D Drawing PID-1 -WLD-LR20221License Renewal Pl D Drawing Pl D- 1 -VSL-LR2O77 6License Renewal PID Drawing PID-1-CBS-1R20233License Renewal PID Drawing PID-1-CS-LR20722License Renewal PID Drawing PID-1-CS-LR20725License Renewal PID Drawing PID-1-RC-LR20841License Renewal PID Drawing PID-1-RC-LR20844License Renewal PID Drawing PID-1-RH-1R20662Corrective Action Documents19849595-3370598-0080498-0166199-1256200-0528601-0420401-0437301-0741701-0875101-0877001-0238901-1342902-01 98902-0221102-0313202-0511202-0569802-0867002-0867102-1342502-1517702-1702703-0353603-0741804-1 138904-1263105-0476805-0507805-0754805-0773005-0983205-1 305605-1509305-041 1 506-0885506-1112107-0374107-0514407-0937707-1235607-1415807-1 559907-14047Attachment
A-608-0579508-0603308-0608008-0608808-1 31 7308-0146108-0146808-1370608-1527709-0148909-01 52009-20735200-21696800-59082401-63276Apparent Cause Evaluation for B EDG rocker arm lube oil tank fuel dilutionApparent Cause Evaluation for supply jug oil contamination with waterApparent Cause Evaluation for aluminum bronze fittings in sea water piping systemsMiscellaneous09CAR029, Change Authorization Request: De-Watering System for Safety Related CableVaults, Dated 6/25109Keyword searches of CRs for Karl Fischer, water contamination, cast iron, graphitization,dezincification, de-alloy, and leachingFire Protection System Walk Down ReportPlant Engineering Guidelines System Walkdowns PEG-10 Revision 19Roving NSO Log Operations Routine Tours, 0210912011Buried Piping Program ER-AA-102Buried Piping Program ER-AA-1 02-1000Mechanical Maintenance Procedure "Application of Repair and Protective Coating(s)"MS0517.12 Rev. 04, Chg. 03Svstem Health ReportsSystem Health Reports, Switchyard System, Dated 111109 through 12131110Cable Program Health Report, Dated 1011log through 12131110Predictive Maintenance Program Health Report, Quarter 4,2007 to Quarter 3, 2008Predictive Maintenance Program Health Report, Quarter 4,2OOg to Quarter 2,2010Buried Piping Program Health Report - 4n Quarter 2008 through 3'o Quarter 2010Cathodic Protection System Health Report 1't Quarter 2004 through 3'o Quarter 2010Above Ground Steel Tanks Program Health Report 1010112008 - 12/3112008Above Ground Steel Tanks Program Health Report 0110112009 - 03/3112009Above Ground SteelTanks Program Health Report 0410112009 - 06/30/2009Above Ground Steel Tanks Program Health Report 0710112009 - 09/30/2009Above Ground Steel Tanks Program Health Report 10/01/2009 - 1213112009Above Ground Steel Tanks Program Health Report 0110112010 - 0313112010Above Ground SteelTanks Program Health Report 0410112010 - 06/30/2010Attachment
A-7Above Ground SteelTanks Program Health Report 0710112010 - 09/30/2010Above Ground Steel Tanks Program Health Report 10lO1l201A - 1213112010RHR System Health Report 1UA112010 - 1213112010RHR System Health Report 2010-04RHR System Walk-Down Report 0210812011RHR System Walk-Down Report 0410112010RHR System Walk-Down Report 06/30/2010Attachment
 
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