Text

South Texas Project, Units 1 and 2 - 2013 Annual Update to License Renewal Application
	ML13310A311
Person / Time
Site:	South Texas
Issue date:	10/28/2013
From:	Powell G T; South Texas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-1 3003041, STI: 33764926, TAC ME4936, TAC ME4937
	Download: ML13310A311 (37)
	v • d • e

Nuclear Operating CompanySouth exas Pro/ect Electric Generating Station PO Box 289 Wadsworth, Te-as 77483 AX A AOctober 28, 2013NOC-AE-1 300304110 CFR 54STI: 33764926U. S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555-0001South Texas ProjectUnits 1 and 2Docket Nos. STN 50-498, STN 50-4992013 Annual Update to the South Texas ProjectLicense Renewal Application (TAC NOS. ME4936 and ME4937)

Reference:

STPNOC Letter dated October 25, 2010, from G. T. Powell to NRC DocumentControl Desk, "License Renewal Application" (NOC-AE-10002607) (ML103010257)By the referenced letter, STP Nuclear Operating Company (STPNOC) submitted an applicationto the Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating LicensesNPF-76 and NPF-80, for South Texas Project (STP) Units 1 and 2, respectively. Theapplication included the License Renewal Application (LRA), and the Applicant's EnvironmentalReport -Operating License Renewal Stage. As required by 10 CFR 54.21(b), each yearfollowing submittal of the LRA, an amendment to the LRA must be submitted that identifies anychange to the current licensing basis (CLB) that materially affects the contents of the LRA,including the Updated Final Safety Analysis Report (UFSAR) supplement.This LRA update covers the period from September 1, 2012 through August 31, 2013.Enclosure 1 identifies STP LRA changes that are being made to: (1) reflect the CLB thatmaterially affect the LRA; and (2) reflect completed enhancements and commitments. Changesto LRA pages described in Enclosure 1 are depicted as line-in/line-out pages provided inEnclosure 2.License Renewal Application revised regulatory commitments are provided in Enclosure 2.There are no other regulatory commitments in this letter.Should you have any questions regarding this letter, please contact Arden Aldridge, STPLicense Renewal Project Lead, at (361) 972-8243, or Ken Taplett, STP License RenewalProject regulatory point-of-contact, at (361) 972-8416.

NOC-AE-1 3003041Page 2I declare under penalty of perjury that the foregoing is true and correct.Executed on OoP,66 z, zol3DateG. T. PowellSite Vice PresidentKJT

Enclosures:

1. STPNOC License Renewal Application (LRA) Changes Reflected in 2013Annual LRA Update2. STP LRA Changes with Line-in/Line-out Annotations NOC-AE-13003041Page 3cc:(paper copy)(electronic copy)Regional Administrator, Region IVU. S. Nuclear Regulatory Commission1600 East Lamar BoulevardArlington, Texas 76011-4511Balwant K. SingalSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North (MS 8B1)11555 Rockville PikeRockville, MD 20852NRC Resident InspectorU. S. Nuclear Regulatory CommissionP. O. Box 289, Mail Code: MN116Wadsworth, TX 77483Jim CollinsCity of AustinElectric Utility Department721 Barton Springs RoadAustin, TX 78704John W. DailyLicense Renewal Project Manager (Safety)U.S. Nuclear Regulatory CommissionOne White Flint North (MS 011-Fl)Washington, DC 20555-0001Tam TranLicense Renewal Project Manager(Environmental)U. S. Nuclear Regulatory CommissionOne White Flint North (MS 011 F01)Washington, DC 20555-0001A. H. Gutterman, EsquireKathryn M. Sutton, EsquireMorgan, Lewis & Bockius, LLPJohn RaganChris O'HaraJim von SuskilNRG South Texas LPKevin PolioRichard PenaCity Public ServicePeter NemethCrain Caton & James, P.C.C. MeleCity of AustinRichard A. RatliffRobert FreeTexas Department of State Health ServicesBalwant K. SingalJohn W. DailyTam TranU. S. Nuclear Regulatory Commission NOC-AE-13003041Enclosure 1STPNOC License Renewal Application (LRA)Changes Reflected in2013 Annual LRA Update NOC-AE-13003041Page 1 of 2STPNOC License Renewal Application (LRA)Changes Reflected in 2013 Annual LRA UpdateFollowing Changes Materially Affect the LRAReason for Change Affected LRASections or TablesRevised Section 2.1.2.3.5, "Station Blackout" to reflect 2.1.2.3.5Change Notice 3013 to the Updated Final Safety AnalysisReport (UFSAR) that deleted the section on "StationBlackout Coping Duration". (Note 1)Revised Final Safety Analysis Report Supplement to reflect A1.35that the PWR Reactor Internals program is a new programthat has now been implemented.South Texas Project UFSAR Change Notice 3073 deleted a A1.37discussion in Appendix 9A of the UFSAR that relief requestsfor leaks in piping diameters of one inch or less areexempted per the ASME code. The aging managementprogram was revised to delete any indication thatcomponents with indications of through-wall dealloying,associated with piping of one inch in diameter or less, wouldnot be replaced by the end of the next refueling outage.Clarify that procedures will be enhanced to perform a remote B2.1.3VT-1 of stud insert #30 "in Unit 2 only" concurrent with thevolumetric examination once every 10 years to verify noadditional loss of bearing surface area.Revised program description to reflect that the PWR Reactor B2.1.35Internals program is a new program that has now beenimplemented. _South Texas Project UFSAR Change Notice 3073 deleted a B2.1.37discussion in Appendix 9A of the UFSAR that relief requestsfor leaks in piping diameters of one inch or less areexempted per the ASME code. Aging management programrevised to delete any indication that components withindications of through-wall dealloying, associated with pipingof one inch in diameter or less, would not be replaced by theend of the next refueling outage.Updated implementation schedule for License Renewal Table A4-1Commitment Item #27 to implement the PWR ReactorInternals Program as described in MRP-227A to reflect thatthe commitment has been completed.Revised Commitment Item #37 to take groundwater samples Table A4-1to reflect that the commitment has been completed.Revised Commitment Item #42 to reflect that the remote VT- Table A4-11 inspection of stud insert #30 is "in Unit 2 only".Revised Commitment Item #43 to reflect that the Table A4-1commitment to remove seal cap enclosures for the valvesdescribed in Unit 1 has been completed. NOC-AE-13003041Page 2 of 2Note 1: Change notice 3013 to the Updated Final Safety Analysis Report deleted the section on"Station Blackout Coping Duration" from Section 8.3.4 "Station Blackout". STP established andthe NRC acknowledged in a Safety Evaluation that the AAC power source meets therequirements of 10 CFR 50.63 and has sufficient capacity and capability to provide power to theSBO loads within 10 minutes of the onset of SBO. Therefore, the SBO Rule is satisfied for notrequiring a coping analysis. NOC-AE-13003041Enclosure 2STP LRA Changes with Line-in/Line-out Annotations NOC-AE-13003041Page 1 of 302.1.2.3.5 Station BlackoutCriterion 10 CFR 54.4(a)(3) requires that plant SSCs within the scope of license renewalinclude all SSCs relied on in safety analyses or plant evaluations to perform a functionthat demonstrates compliance with the regulations for SBO (10 CFR 50.63).STPNOC provided the NRC with its original response to the SBO Rule incorrespondence dated April 17, 1989. The NRC issued its initial Safety Evaluation onJuly 17, 1991. On July 28, 1994, a STP self-assessment determined that the STP unitswere not fully meeting the requirements of the SBO Rule. On August 1, 1994 STPNOCprepared a justification for continued operation which implemented the 73 miles per hourhurricane shutdown criterion, reduced the coping duration from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , andeliminated crediting the auxiliary ESF transformers for SBO coping. On March 1, 1995STP provided the NRC a revised position to the SBO Rule. The NRC issued a SafetyEvaluation Report for the Revised Station Blackout Position on July 24, 1995.UFSAR Section 8.3.4 discusses SBO coping duration, alternate AC power source,condens-ate inventory' for decay heat removal, class, 1 E. battery' capacity, compressed airrequi'8remets, effect Of loss6 Of ventiation, contaRin.ment isolationR, reactor coolantien~eGy and quality assurance program requirements.The STP SBO analysis was performed using the guidance provided in NUMARC 87-00,Rev. 0 and the coping time (the postulated maximum SBO duration) was determined tobe four hours. The STP SBO position credits any one of the three standby dieselgenerators as the AAC source. Each standby diesel generator can energize anindependent train of auxiliary feedwater, essential cooling water, component coolingwater, steam generator power operated relief valves, high head safety injection, andEAB/Control Room HVAC. Each standby diesel generator meets or exceeds theNUMARC 87-00, Appendix B, criteria for capacity, capability and connectability.STP's offsite power system is in accordance with GDC 17 and provides two separatepaths of power from the transmission system to the ESF buses as shown onFigure 2.1-2. Recovery from an SBO focuses on restoration of an AC power source.This can be from the onsite diesel generators, or from an offsite source. STP has thefollowing paths of offsite power.* 345 kV switchyard to main and unit auxiliary transformers* 345 kV switchyard to standby transformer 1* 345 kV switchyard to standby transformer 2The main and unit auxiliary transformers are connected to the switchyard throughdisconnect G01 9 (Unit 1), and G029 (Unit 2) which connects to the switchyard viaswitchyard circuit breakers Y510 and Y520 (Unit 1), and Y590 and Y600 (Unit 2). Theunit auxiliary transformer, the iso-phase bus, the main transformer, the overheadtransmission lines, the switchyard breakers and switchyard breaker control cables andconnections are within the scope of license renewal.Standby transformers 1 and 2 are connected to the 345 kV switchyard north (Unit 1) andsouth (Unit 2) bus via disconnect S014 (Unit 1), and S024 (Unit 2). The standbytransformers, the overhead transmission lines, and the switchyard disconnects are within NOC-AE-1 3003041Page 2 of 30the scope of license renewal.A position paper was created to summarize the results of a review of the SBOdocumentation for STP. The position paper identifies the SSCs credited with coping andrecovering from a SBO. The SSCs identified in the SBO position paper were used inscoping evaluations to identify SSCs that demonstrate compliance with 10 CFR 50.63.License renewal drawing LR-STP-ELEC-OOOOOEOAAAA schematically shows theportions of the plant AC electrical distribution system, including the SBO recovery path,that are included within the scope of license renewal and is summarized in Figure 2.1-2,Station Blackout Recovery Path.SSCs classified as satisfying criterion 10 CFR 54.4(a)(3) related to station blackout areidentified as within the scope of license renewal. NOC-AE-13003041Page 3 of 30A1.35 PWR REACTOR INTERNALSThe PWR Reactor Internals program manages cracking, loss of material, loss of fracturetoughness, dimensional changes, and loss of preload for reactor vessel components thatprovide a core structural support intended function. The program implements theguidance of EPRI 1022863, PWR Internals Inspection and Evaluation Guideline (MRP-227-A) and EPRI 1016609, Inspection Standard for PWR Internals (MRP-228). Theprogram manages aging consistent with the inspection guidance for Westinghousedesignated primary components in Table 4-3 of MRP-227-A and Westinghousedesignated expansion components in Table 4-6 of MRP-227-A. The expansioncomponents are specified to expand the primary component sample should theindications of the sample be more severe than anticipated. The aging effects of a thirdset of MRP-227-A internals locations are deemed to be adequately managed by existingprogram components whose aging is managed consistent with ASME Section XI TableIWB-2500-1, Examination Category B-N-3.Program examination methods include visual examination (VT-3), enhanced visualexamination (EVT-1), volumetric examination, and physical measurements. Theprogram provides both examination acceptance criteria for conditions detected as aresult of monitoring the primary components, as well as criteria for expandingexaminations to the expansion components when warranted by the level of degradationdetected in the primary components. Based on the identified aging effect, andsupplemental examinations if required, the disposition process results in an evaluationand determination of whether to accept the condition until the next examination orimplement corrective actions. Any detected conditions that do not satisfy theexamination acceptance criteria are required to be dispositioned through the correctiveaction program, which may require repair, replacement, or analytical evaluation forcontinued service until the next inspection.The PWR Reactor Internals program is a new program and has been will-beimplemented within 24 months aftor the issua-nce f EPRI 1022863, PWR InternplsIn ....ton And Ev'luation Gu-ideine MRP 227 A. NOC-AE-13003041Page 4 of 30A1.37 Selective Leaching of Aluminum BronzeThe Selective Leaching of Aluminum Bronze program manages loss of material due toselective leaching of aluminum bronze (copper alloy with greater than eight percentaluminum) components exposed to raw water within the scope of license renewal. TheSelective Leaching of Aluminum Bronze program is an existing program that isimplemented by STP procedure. The procedure directs that every six months (not toexceed nine months), an inspection of all aluminum bronze components be completed.STP has buried piping with less than eight percent aluminum content, and that is notsusceptible to dealloying. However, there are welds in which the filler metal is a copperalloy with greater than eight percent aluminum material. Therefore, the proceduredirects that a yard walkdown be performed above the buried piping with aluminumbronze welds, from the intake structure to the unit and from the unit to the dischargestructure to look for changes in ground conditions that would indicate leakage. If a leakfrom below-grade weld is discovered by surface water monitoring or during a buriedECW piping inspection, a section of each leaking weld will be removed for destructiveexamination. Aluminum bronze (copper alloy with greater than 8 percent aluminum)components which are found to have indications of through-wall dealloying areevaluated, and scheduled for replacement by the corrective action program.Components with indications of through-wall dealloying, ascociatod with Piping gratertthaR Gon inch in diameter, will be replaced by the end of the next refueling outage.Volumetric examinations of aluminum bronze material components that demonstrateexternal leakage will be performed where the configuration supports this type ofexamination to conclude with reasonable assurance that cracks are not approaching acritical size.Destructive examination of each leaking component removed from service will beperformed to determine the degree of dealloying until 10 percent of the susceptiblecomponents in the ECW system are examined. The degree of dealloying and crackingwill be trended by comparing examination results with previous examination results.Metallurgical testing of leaking aluminum bronze material components in the ECWsystem removed from service will be performed to update the structural integrityanalyses, to confirm load carrying capacity and to determine the degree of dealloying bydestructive examination. Metallurgical testing of the removed leaking component will beperformed until at least three different size components of two samples each are tested,and at least nine total samples are tested. The metallurgical testing will include fracturetoughness testing of test samples that include a crack in the dealloyed material wheresufficient sample size supports bend testing. Additionally, the samples will be tested forchemical composition including aluminum content, mechanical properties (such as yieldand ultimate tensile strengths) and microstructure. Ultimate tensile strength will betrended and compared to the acceptance criterion. The degree of dealloying andcracking will be trended by comparing examination results with previous examinationresults. NOC-AE-1 3003041Page 5 of 30Beginning 10 years prior to the period of extended operation for each 10-year interval,periodic metallurgical testing of aluminum bronze material components will be performedto update the structural integrity analyses, confirm load carrying capacity, and determinedegree of dealloying. For each 10-year interval beginning 10 years prior to the period ofextended operation, 20 percent of leaking above ground components removed fromservice, but at least one, will be tested every five years. Tensile test samples from aremoved component will be tested to include both leaking and non-leaking portions ofthe component. If at least two leaking components are not identified two years prior tothe end of each 10-year testing interval, a risk-ranked approach based on thosecomponents most susceptible to degradation will be used to identify candidatecomponents for removal and testing so at least two components are tested during the10-year interval. The samples will be tested for chemical composition includingaluminum content, mechanical properties (such as yield and ultimate tensile strengths)and microstructure. The samples will be destructively examined to determine the degreeof dealloying and the presence of cracks. Ultimate tensile strength will be trended andcompared to the acceptance criterion. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.An engineering evaluation will be performed at the end of each test to determine if thesample size requires adjustment based on the results of the tests.The acceptance criterion for ultimate tensile strength value of aluminum bronze materialis greater than or equal to 30 ksi. The acceptance criterion for the fracture toughness isgreater than or equal to 65 ksi in"12 for aluminum bronze castings and at welded joints inthe heat affected zones. The acceptance criterion for yield strength is equal to orgreater than one-half of the ultimate strength. If a criterion is not met, the condition willbe documented in the corrective action program to perform a structural integrity analysisto confirm that the load carrying capacity of the tested material remains adequate tosupport the intended function of the ECW system through the period of extendedoperation. NOC-AE-1 3003041Page 6 of 30B2.1.3Reactor Head Closure StudsProgram DescriptionThe Reactor Head Closure Studs program manages cracking and loss of material byconducting ASME Section XI inspections of reactor vessel flange stud hole threads,reactor head closure studs, nuts, washers, and bushings. The program includesperiodic visual, surface, and volumetric examinations of reactor vessel flange stud holethreads, reactor head closure studs, nuts, washers, and bushings and performs visualinspections of the reactor vessel flange closure during primary system leakage tests.The STP program implements ASME Section X1 code, Subsection IWB, 2004 Edition.Reactor vessel flange stud hole threads, reactor head closure studs, nuts, washers, andbushings are identified in ASME Section XI Tables IWB-2500-1 and are within the scopeof license renewal. The program implements recommendations in NUREG-1 339 andNRC Regulatory Guide 1.65, Material and Inspection for Reactor Vessel Closure Studs,to address reactor head stud bolting degradation except for yield strength of existingbolting materials. STP uses lubricants on reactor head closure stud threads after reactorhead closure stud, nut, and washer cleaning and examinations are complete. Thelubricants are compatible with the stud material and operating environment and do notinclude MoS2 which is a potential contributor to stress corrosion cracking.In conformance with 10 CFR 50.55a(g)(4)(ii), the STP ISI Program is updated duringeach successive 120-month inspection interval to comply with the requirements of thelatest edition of the Code specified twelve months before the start of the inspectioninterval. STP will use the ASME Code Edition consistent with the provisions of10 CFR 50.55a during the period of extended operation.Potential cracking and loss of material conditions in reactor vessel flange stud holethreads, reactor head closure studs, nuts, washers, and bushings are detected throughvisual, surface, or volumetric examinations in accordance with ASME Section X1requirements in STP procedures every ten years. A remote VT-1 of stud insert #30 (Unit2 only) is performed concurrent with the volumetric examination once every 10 years toverify no additional loss of bearing surface area. These inspections are conducted duringrefueling outages. Reactor vessel studs are removed from the reactor vessel flangeeach refueling outage. Studs, nuts, washers, and bushings are stored in protectiveracks after removal. Reactor vessel flange holes are plugged with water tight plugsduring cavity flooding. These methods assure the holes, studs, nuts, washers, andbushings are protected from borated water during cavity flooding. Reactor vessel flangeleakage is detected prior to reactor startup during reactor coolant system pressuretesting each refueling outage. The STP program has proven to be effective inpreventing and detecting potential aging effects of reactor vessel flange stud holethreads, closure studs, nuts, washers, and bushings.NUREG-1801 ConsistencyThe Reactor Head Closure Studs program is an existing program that is consistent, withexception to NUREG-1801,Section XI.M3, Reactor Head Closure Studs. NOC-AE-13003041Page 7 of 30Exceptions to NUREG-1 801Program Elements Affected:Scope of Program (Element 1)Regulatory Guide 1.65 states that the ultimate tensile strength of stud bolting materialshould not exceed 170 ksi. One closure head insert has a tensile strength of 174.5 ksi.STP credits inservice inspections that are within the scope of this AMP, which areimplemented in accordance with the STP Inservice Inspection Program, ExaminationCategory B-G-1 requirements, as the basis for managing cracking in these components.This is in accordance with the "parameters monitored or inspected" and "detection ofaging effects" program elements in NUREG 1801,Section XI.M3. In addition, the studs,nuts and washers are coated with a lubricant which is compatible with the stud materials,and the studs, nuts, and washers are protected from exposure to boric acid by removingthem and plugging the reactor vessel flange holes during cavity flooding. Replacementreactor head closure bolting obtained in the future (not currently installed or on site asspare parts) will be fabricated from material with an actual measured yield strength lessthan 150 ksi.Corrective Actions (Element 7)NUREG-1801,Section XI.M3 specifies the use of Regulatory Guide 1.65 requirementsfor closure stud and nut material. STP uses SA-540, Grade B-24 (as modified by CodeCase 1605) stud material. The use of this material has been found acceptable to theNRC for this application within the limitations discussed in Regulatory Guide 1.85,Materials Code Case Acceptability.EnhancementsScope of Program (Element 1)Procedures will be enhanced to preclude the future use of replacement closure studassemblies fabricated from material with an actual measured yield strength greater thanor equal to 150 ksi. The use of currently installed components and any sparecomponents currently on site is allowed.Detection of Aging Effects (Element 4)Procedures will be enhanced to perform a remote VT-1 of stud insert #30 (Unit 2 only)concurrent with the volumetric examination once every 10 years to verify no additionalloss of bearing surface area.Operating ExperienceReview of plant-specific operating experience has not revealed any program adequacyissues with the Reactor Head Closure Studs program for reactor vessel closure studs,nuts, washers, bushings, and flange thread holes. No cases of cracking due to SCC orIGSCC have been identified with STP reactor vessel studs, nuts, washers, bushings,and flange stud holes. NOC-AE-13003041Page 8 of 30Review of the Refueling Outage Inservice Inspection Summary Reports for Interval 2indicates there were no repair/replacement items identified with reactor vessel closurestuds, nuts, washers, bushings, or flange thread holes. None of the repair/replacementitems indicate any implementation issues with the STP ASME Section XI Program forreactor closure studs, nuts, washers, bushings, or flange thread holes.The ISI Program at STP is updated to account for industry operating experience. ASMESection XI is also revised every three years and addenda issued in the interim, whichallows the code to be updated to reflect operating experience. The requirement toupdate the ISI Program to reference more recent editions of ASME Section XI at the endof each inspection interval ensures the ISI Program reflects enhancements due tooperating experience that have been incorporated into ASME Section XI.ConclusionThe continued implementation of the Reactor Head Closure Studs program providesreasonable assurance that aging effects will be managed such that the systems andcomponents within the scope of this program will continue to perform their intendedfunctions consistent with the current licensing basis for the period of extended operation. NOC-AE-13003041Page 9 of 30B2.1.35 PWR Reactor InternalsProgram DescriptionThe PWR Reactor Internals program manages cracking, loss of material, loss of fracturetoughness, dimensional changes, and loss of preload for reactor vessel components thatprovide a core structural support intended function. The program implements theguidance of EPRI 1022863, PWR Internals Inspection and Evaluation Guideline (MRP-227-A) and EPRI 1016609, Inspection Standard for PWR Internals (MRP-228, Rev. 0).The program manages aging consistent with the inspection guidance for Westinghousedesignated primary components in Table 4-3 of MRP-227-A, Westinghouse designatedexpansion components in Table 4-6 of MRP-227-A, and the Westinghouse designatedexisting components in Table 4-9 of MRP-227-A. Primary components are expected toshow the leading indications of the degradation effects. The expansion components arespecified to expand the primary component sample should the indications of the samplebe more severe than anticipated. The aging effects of a third set of MRP-227-A internalslocations are deemed to be adequately managed by existing program componentswhose aging is managed consistent with ASME Section XI Table IWB-2500-1,Examination Category B-N-3.Program examination methods include visual examination (VT-3), enhanced visualexamination (EVT-1), volumetric examination, and physical measurements. Boltingultrasonic examination technical justifications in MRP-228 have demonstrated theindication detection capability to detect loss of integrity of PWR internals bolts, pins, andfasteners, such as baffle-former bolting. For some components, the MRP-227-Amethodology specifies a focused visual (VT-3) examination, similar to the current ASMECode,Section XI, Examination Category B-N-3 examinations, in order to determine thegeneral mechanical and structural condition of the internals by (a) verifying parameters,such as clearances, settings, and physical displacements; and (b) detectingdiscontinuities and imperfections, such as loss of integrity at bolted or weldedconnections, loose or missing parts, debris, corrosion, wear, or erosion. In some cases,VT-3 visual methods are used for the detection of surface cracking when the componentmaterial has been shown to be tolerant of easily detected large flaws. In some cases,where even more stringent examinations are required, enhanced visual (EVT-1)examinations or ultrasonic methods of volumetric inspection, are specified for certainselected components and locations.The program provides both examination acceptance criteria for conditions detected as aresult of monitoring the primary components, as well as criteria for expandingexaminations to the expansion components when warranted by the level of degradationdetected in the primary components. Based on the identified aging effect, andsupplemental examinations if required, the disposition process results in an evaluationand determination of whether to accept the condition until the next examination orimplement corrective actions. Any detected conditions that do not satisfy theexamination acceptance criteria are required to be dispositioned through the correctiveaction program, which may require repair, replacement, or analytical evaluation forcontinued service until the next inspection. NOC-AE-13003041Page 10 of 30The PWR Vessel Internals program is a new program that has beenwil-be implementedwfithin 24 monthr, after the icuneof MIRP 227 A, PWVR IntFRnasInseto nEvalua-tion Guidelino. The program will include future industry operating experience, asit is incorporated into the future revisions of MRP-227-A, to provide reasonableassurance for long-term integrity of the reactor internals. The reactor vessel internalsincluded in the scope of the PWR Reactor Internals program are identified in Element 1.The scope of the program does not include welded attachments to the internal surface ofthe reactor vessel because these components are managed by the ASME Section XlInservice Inspection, Subsections IWB, IWC, and IWD program (B2.1.1) (exam categoryB-N-2) and /or the Nickel-Alloy Aging Management Program (B2.1 .34). The scope of theprogram also does not include BMI flux thimble tubes which are managed by the FluxThimble Tube Inspection program (B2.1.21).Aging Management Program ElementsThe results of an evaluation of each element against the 10 elements described inAppendix A of NUREG-1800, Standard Review Plan for Review of License RenewalApplications for Nuclear Power Plants are provided below.Scope of Program -Element 1The scope of the program applies the guidance in MRP-227-A which providesaugmented inspection and flaw evaluation methodology for assuring the functionalintegrity of Westinghouse reactor vessel internals. The scope of the PWR ReactorInternals program includes components that provide a core structural support intendedfunction and are managed by the Westinghouse designated primary components inTable 4-3 of MRP-227-A and Westinghouse designated expansion components in Table4-6 of MRP-227-A and applicable MRP-227-A methodology license renewal applicantaction items. MRP-227-A Table 4-9 also identifies existing program components whoseaging is managed consistent with ASME Section XI Table IWB-2500-1, ExaminationCategory B-N-3.Primary components are expected to show the leading indications of the degradationeffects. The expansion components are specified to expand the primary componentsample should the indications of the sample be more severe than anticipated. The agingeffects of a third set of MRP-227-A internals locations are deemed to be adequatelymanaged by existing program components whose aging is managed consistent withASME Section Xl Table IWB-2500-1, Examination Category B-N-3.The STP reactor vessel internals are divided into the following major component groups:the lower core support assembly (including the entire core barrel assembly, baffle-formerassembly, neutron shield panel, core support plate, and energy absorber assembly), theupper core support (UCS) assembly (including the upper support plate, support column,control rod guide tube assembly, upper core plate, and protective skirt), the incoreinstrumentation support structures (including the instrumentation columns (exitthermocouples), upper/lower tie plates, and instrumentation columns (BMI)), andmiscellaneous alignment/interface components (including internals hold-down spring,upper core plate guide pins, and radial support keys including clevis inserts). NOC-AE-13003041Page 11 of 30The following reactor vessel internals are included in the scope of the PWR ReactorInternals program:1. Control rod guide tube assembly and Bolting-Guide plate (cards) [Primary component]-Lower flange welds and adjacent base metal (Addressed in AMR by ComponentType of "RVI Control Rod Guide Tube Assembly") [Primary component]-Guide Tube Support Pins (Split Pins) (Addressed in AMR by Component Type of"RVI Control Rod Guide Tube Bolting") [Existing programs component]2. Core barrel assembly-Upper core barrel flange weld and adjacent base metal (Addressed in AMR byComponent Types of "RVI Core Barrel Assembly") [Primary component]-Core barrel assembly-former bolting [Expansion component]Core barrel flange (Addressed in AMR by Component Types of "RVI Core BarrelAssembly") [Expansion component and Existing programs component]-Core barrel axial welds and adjacent base metal [Expansion component]-Core barrel girth welds and adjacent base metal [Primary component]-Core barrel outlet nozzle welds and adjacent base metal [Expansion component]-Lower core barrel flange weld and adjacent base metal (Addressed in AMR byComponent Types of "RVI Core Barrel Assembly") [Primary component]3. Baffle-former assembly and bolting-Baffle-edge bolting [Primary component]-Baffle-former bolting [Primary component]-Baffle-former assembly [Primary component]4. Alignment and interfacing components-Internals hold-down spring [Primary component]-Radial support key clevis insert bolts [Existing programs component]-Upper core plate guide pins [Existing programs component]5. Instrumentation support structures-Instrumentation columns -BMI [Expansion component] NOC-AE-13003041Page 12 of 306. Upper core support assemblyUpper core support protective skirt [Existing programs component]Upper Core Plate [Expansion component]7. Lower Core Support StructureCore Support Plate Forging [Expansion component]The scope of the program also does not include welded attachments to the internalsurface of the reactor vessel because these components are managed by the ASMESection XI Inservice Inspection, Subsections IWB, IWC, and IWD program (B2.1.1)(exam category B-N-2) and /or the Nickel-Alloy Aging Management Program (B2.1.34).The scope of the program also does not include BMI flux thimble tubes which aremanaged by the Flux Thimble Tube Inspection program (B2.1.21).The STP reactor vessel internals configuration does not include the lower internalsassembly (lower support column bodies and lower core plate) noted in MRP-227-A.The PWR Reactor Internals program is consistent with the following MRP-227-Aassumptions (determination of applicability) which are based on PWR representativeinternals configurations and operational histories.(1) STP has operated for less than 30 years of operation with high leakage core loadingpatterns. Operation with high leakage core loading was followed by implementationof a low-leakage fuel management pattern for the remaining operating life.(2) STP operates at fixed power levels and does not usually vary power based oncalendar or load demand schedule.(3) STP has not implemented any design changes beyond those identified in industryguidance or recommended by Westinghouse.Preventive Actions -Element 2The PWR Reactor Internals program does not prevent degradation due to aging effects,but provides measures for monitoring to detect the degradation prior to loss of intendedfunction. Preventive measures to mitigate aging effects such as loss of material andcracking include monitoring and maintaining reactor coolant water chemistry consistentwith the guidelines of EPRI TR 1014986, PWR Primary Water Chemistry Guidelines,Volume 1. The primary water chemistry program is described separately in the WaterChemistry program (B2.1.2).Parameters Monitored or Inspected -Element 3The PWR Reactor Internals program monitors the following aging effects by inspectionin accordance with the guidance of MRP-227-A or ASME Section XI Category B-N-3:(1). CrackingCracking is due to stress corrosion cracking (SCC), primary water stress corrosion NOC-AE-13003041Page 13 of 30cracking (PWSCC), irradiation assisted stress corrosion cracking (IASCC), or fatigue/cyclical loading. Cracking is monitored with a visual inspection for evidence of surfacebreaking linear discontinuities or a volumetric examination. Surface examinations mayalso be used to supplement visual examinations for detection and sizing of surface-breaking discontinuities.(2). Loss of MaterialLoss of Material is due to wear. Loss of material is monitored with a visual inspection forgross or abnormal surface conditions.(3). Loss of Fracture ToughnessLoss of fracture toughness is due to thermal aging or neutron irradiation embrittlement.The impact of loss of fracture toughness is indirectly monitored by using visual orvolumetric examination techniques to monitor for cracking and by applying applicablereduced fracture toughness properties in the flaw evaluations if cracking is detected andis extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.(4). Dimensional ChangesDimensional Changes are due to void swelling and irradiation growth, distortion ordeflection. The program supplements visual inspection with physical measurements tomonitor for any dimensional changes due to void swelling, irradiation growth, distortion,or deflection.(5). Loss of PreloadLoss of preload is caused by thermal and irradiation-enhanced stress relaxation orcreep. Loss of preload is monitored with a visual inspection for gross surface conditionsthat may be indicative of loosening in applicable bolted, fastened, keyed, or pinnedconnections.The PWR Reactor Internals program manages the aging effects noted above consistentwith the inspection guidance for Westinghouse designated primary components in Table4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 ofMRP-227-A. MRP-227-A also identifies Existing Program components whose aging ismanaged consistent with ASME Section XI Table IWB-2500-1, Examination Category B-N-3. See the component list in element 1 to identify Primary, Expansion, and Existingcomponents.Detection of Aging Effects -Element 4The PWR Reactor Internals program detects aging effects through the implementation ofthe parameters monitored or inspected criteria and bases for Westinghouse designatedPrimary Components in Table 4-3 of MRP-227-A and for Westinghouse designatedExpansion Components in Table 4-6 of MRP-227-A. The aging effects of a third set ofMRP-227-A internals locations identified in Table 4-9 of MRP-227-A are deemed to beadequately managed by existing program components whose aging is managedconsistent with ASME Section XI Table IWB-2500-1, Examination Category B-N-3.One hundred percent of the accessible volume/area of each component will beexamined for the Primary and Expansion components inspection category components.The minimum examination coverage for primary and expansion inspection categories is75 percent of the component's total (accessible plus inaccessible) inspection NOC-AE-1 3003041Page 14 of 30area/volume be examined. When addressing a set of like components (e.g. bolting), theminimum examination coverage for primary and expansion inspection categories is 75percent of the component's total population of like components (accessible plusinaccessible).If defects are discovered during the examination, STP enters the information into theSTP corrective action program and evaluates whether the results of the examinationensure that the component (or set of components) will continue to meet its intendedfunction under all licensing basis conditions of operation until the next scheduledexamination. Engineering evaluations that demonstrate the acceptability of a detectedcondition will be performed consistent with WCAP-17096-NP.Monitoring and Trending -Element 5The program provides both examination acceptance criteria (See Element 6) forconditions detected as a result of monitoring the primary components as described inElement 4, as well as criteria for expanding examinations to the expansion componentswhen warranted by the level of degradation detected in the primary components. Basedon the identified aging effect, and supplemental examinations if required, the dispositionprocess results in an evaluation and determination of whether to accept the conditionuntil the next examination or implement corrective actions. Any detected conditions thatdo not satisfy the examination acceptance criteria (See Element 6) are required to bedispositioned through the corrective action program (See Element 7), which may requirerepair, replacement, or analytical evaluation for continued service until the nextinspection.Acceptance Criteria -Element 6Examination acceptance for the Primary and Expansion component examinations areconsistent with Section 5 of MRP-227-A. ASME Section XI section IWB-3500acceptance criteria apply to Existing Programs components. The following examinationacceptance criteria apply to the STP reactor vessel internals:Visual examination (VT-3) and enhanced visual examination (EVT-1)For existing program components, the ASME Code Section XI, Examination Category B-N-3 provides the following general relevant conditions for the visual (VT-3) examinationof removable core support structures.(1) Structural distortion or displacement of parts to the extent that component functionmay be impaired,(2) Loose, missing, cracked, or fractured parts, bolting, or fasteners,(3) Corrosion or erosion that reduces the nominal section thickness by more than 5percent,(4) Wear of mating surfaces that may lead to loss of function; and NOC-AE-13003041Page 15 of 30(5) Structural degradation of interior attachments such that the original cross-sectionalarea is reduced more than 5 percent.In addition, for the visual examinations (VT-3) of Primary and Expansion components,the PWR Reactor Internals program is consistent with the more specific descriptions ofrelevant conditions provided in Table 5-3 of MRP-227-A. EVT-1 examinations are usedfor detecting small surface breaking cracks and surface crack length sizing when used inconjunction with sizing aids. EVT- 1 examination has been selected to be theappropriate NDE method for detection of cracking in plates or their welded joints. Therelevant condition applied for EVT-1 examination is the same as found for cracking inASME Section XI section 3500 which is crack-like surface breaking indications.Volumetric examinationIndividual bolts are accepted (pass/fail acceptance) based on the detection of relevantindications established as part of the examination technical justification. When arelevant indication is detected in the cross-sectional area of the bolt, it is assumed to benon-functional and the indication is recorded. Bolted assemblies are evaluated foracceptance based on meeting a specified number and distribution of functional bolts.Acceptance criteria for volumetric examination of STP reactor internals bolting areconsistent with Table 5-3 of MRP-227-A.Physical MeasurementsContinued functionality of the internals hold down spring is confirmed by direct physicalmeasurement. The examination acceptance criterion for this measurement is consistentwith Table 5-3 of MRP-227-A and requires that the remaining compressible height of thespring shall provide hold-down forces within the plant-specific design tolerance.Corrective Actions -Element 7The following corrective actions are available for the disposition of detected conditionsthat exceed the examination acceptance criteria:(1) Supplemental examinations to further characterize and potentially dispose of adetected condition consistent with Section 5.0 of MRP-227-A;(2) Engineering evaluation that demonstrates the acceptability of a detected conditionconsistent with WCAP-17096-NP;(3) Repair, in order to restore a component with a detected condition to acceptablestatus (ASME Section XI); or(4) Replacement of a component with an unacceptable detected condition (ASMESection XI)(5) Other alternative corrective action bases if previously approved or endorsed by theNRC.Relevant indications failing to meet applicable acceptance criteria are repaired orreplaced in accordance with plant procedures. Appropriate codes and standards are NOC-AE-13003041Page 16 of 30specified in both the "ASME Section XI Repair, Replacement, and Post-MaintenancePressure Testing" procedure and in design drawings. Quality assurance requirementsfor repair and replacement activities are also included in the STP Operations QualityAssurance Plan.STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and areacceptable in addressing corrective actions. The QA program includes elements ofcorrective action, confirmation process and administrative controls, and is applicable tothe safety-related and non-safety related systems, structures, and components that aresubject to aging management review.Confirmation Process -Element 8STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and areacceptable in addressing the confirmation process. The QA program includes elementsof corrective action, confirmation process and administrative controls and is applicable tothe safety-related and non-safety related systems, structures and components that aresubject to aging management review.Administrative Controls -Element 9STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and areacceptable in addressing administrative controls. The QA program includes elements ofcorrective action, confirmation process and administrative controls and is applicable tothe safety-related and non-safety related systems, structures and components that aresubject to aging management review.Operating Experience -Element 10Relatively few incidents of PWR internals aging degradation have been reported inoperating U.S. commercial PWR plants. However, a considerable amount of PWRinternals aging degradation has been observed in European PWRs, with emphasis oncracking of baffle-former bolting. The experience reviewed includes NRC InformationNotice 84-18, Stress Corrosion Cracking in PWR Systems and NRC Information Notice98-11, Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants. Mostof the industry operating experience reviewed has involved cracking of austeniticstainless steel baffle-former bolts, or SCC of high-strength internals bolting. SCC ofcontrol rod guide tube split pins has also been reported.Several other items with existing or suspected material degradation concerns that havebeen identified for PWR components are wear in thimble tubes and potentially in controlguide cards and observed cracking in some high-strength bolting and in control rodguide tube alignment (split) pins. The latter are conditions that have been correctedprimarily through bolt replacement with less susceptible material and improved control ofpre-load.Based on industry operating experience, STP replaced the Alloy-750 guide tube supportpins (split pins) with strained hardened (cold worked) 316 stainless steel pins during NOC-AE-1 3003041Page 17 of 30Refueling Outage 1 RE1 2 (Spring 2005) for Unit 1 and Refueling Outage 2RE1 1 (Fall2005) for Unit 2. The replacement was conducted to reduce the susceptibility for stresscorrosion cracking in the split pins. There were no cracked Alloy X-750 pins discoveredduring the replacement process.The ASME Code,Section XI, Examination Category B-N-3 examinations of core supportstructures conducted during Refueling Outage 1 RE1 5 (Fall 2009) for Unit 1, andRefueling Outage 2RE14 (Spring 2010) for Unit 2, did not identify any conditions thatrequired repair, replacement or evaluation.The ISI Program portion of the PWR Reactor Internals program at STP is updated toaccount for industry operating experience. ASME Section XI is also revised every threeyears and addenda issued in the interim, which allows the code to be updated to reflectoperating experience. The requirement to update the ISI Program to reference morerecent editions of ASME Section XI at the end of each inspection interval ensures the ISIProgram reflects enhancements due to operating experience that have beenincorporated into ASME Section XI.With exception of the ASME Section ISI portions, the PWR Reactor Internals programwill be a new program and has no direct programmatic history. A key element of theMRP-227-A program is the reporting of aging of reactor vessel components. STP,through its participation in PWR Owners Group and EPRI-MRP activities, will continue tobenefit from the reporting of inspection information and will share its own operatingexperience with the industry through those groups or INPO, as appropriate.As additional Industry and applicable plant-specific operating experience becomeavailable, the OE will be evaluated and appropriately incorporated into the programthrough the STP Corrective Action and Operating Experience Programs. This ongoingreview of OE will continue throughout the period of extended operation, and the resultswill be maintained on site. This process will confirm the effectiveness of this new licenserenewal aging management program by incorporating applicable OE and performing selfassessments of the program.ConclusionThe implementation of the PWR Reactor Internals program provides reasonableassurance that aging effects will be adequately managed such that the systems andcomponents within the scope of this program will continue to perform their intendedfunctions consistent with the current licensing basis for the period of extended operation. NOC-AE-13003041Page 18 of 30B2.1.37 Selective Leaching of Aluminum BronzeProgram DescriptionThe Selective Leaching of Aluminum Bronze program manages loss of material due toselective leaching for aluminum bronze (copper alloy with greater than eight percentaluminum) components exposed to raw water within the scope of license renewal. Thisplant-specific program will use requirements of the Selective Leaching of Materialsprogram (B2.1.17) specifically relating to aluminum bronze components. The selectiveleaching of aluminum bronze is applied in addition to the Open-Cycle Cooling Waterprogram (B2.1.9).The Selective Leaching of Aluminum Bronze program is an existing program that isimplemented by plant procedure. This procedure directs that every six months (not toexceed nine months), an inspection of aluminum bronze (copper alloy with greater thaneight percent aluminum) components be completed. STP has buried copper piping withless than eight percent aluminum content that is not susceptible to dealloying. However,there are welds in which the filler metal is copper alloy with greater than eight percentaluminum material. Therefore, the procedure directs that a yard walkdown be performedabove the buried piping with aluminum bronze welds, from the intake structure to the unitand from the unit to the discharge structure to look for changes in ground conditions thatindicate leakage. Aluminum bronze (copper alloy with greater than 8 percent aluminum)components which are found to have indications of through-wall dealloying areevaluated, and scheduled for replacement by the corrective action program.Components with indications of through-wall dealloying, greater than one inch, will bereplaced by the end of the next refueling outage. Periodic destructive examinations ofaluminum bronze material components will be performed to update the structuralintegrity analyses, confirm load carrying capacity, and determine degree of dealloying.Aging Management Program ElementsThe results of an evaluation of each element against the 10 elements described inAppendix A of NUREG-1800, Standard Review Plan for Review of License RenewalApplications for Nuclear Power Plants are provided below.Scope of Program (Element 1)The Selective Leaching of Aluminum Bronze program manages loss of material due toselective leaching for aluminum bronze (copper alloy with greater than eight percentaluminum) pumps, piping welds and valve bodies exposed to raw water within the scopeof license renewal. These aluminum bronze (copper alloy with greater than eightpercent aluminum) components with raw water internal environments are susceptible toloss of material due to selective leaching (dealloying).STP has analyzed the effects of dealloying and found that the degradation is slow sothat rapid or catastrophic failure is not a consideration. A structural integrity analysis NOC-AE-13003041Page 19 of 30performed when dealloying was first identified confirmed that 100 percent dealloyedaluminum bronze material retains sufficient load carrying capacity. This structuralintegrity analysis determined that the leakage can be detected before the flaw reaches alimiting size that would affect the intended functions of the essential cooling water andessential cooling water screen wash system.Volumetric examinations of aluminum bronze material components that demonstrateexternal leakage will be performed where the configuration supports this type ofexamination to conclude with reasonable assurance that cracks are not approaching acritical size.Destructive examination of each leaking component removed from service will beperformed to determine the degree of dealloying until 10 percent of the susceptiblecomponents in the ECW system are examined. The degree of dealloying and crackingwill be trended by comparing examination results with previous examination results.Metallurgical testing of leaking aluminum bronze material components in the ECWsystem removed from service will be performed to update the structural integrityanalyses, to confirm load carrying capacity and to determine the degree of dealloying bydestructive examination. Metallurgical testing of the removed leaking component will beperformed until at least three different size components of two samples each are tested,and at least nine total samples are tested. The metallurgical testing will include fracturetoughness testing of test samples that include a crack in the dealloyed material wheresufficient sample size supports bend testing. Additionally, the samples will be tested forchemical composition including aluminum content, mechanical properties (such as yieldand ultimate tensile strengths) and microstructure. Ultimate tensile strength will betrended and compared to the acceptance criterion. Degree of dealloying and crackingwill be trended by comparing examination results with previous examination results.As part of the testing described above, six samples from three aluminum bronzecomponents removed from service in 2012 will be tested for chemical compositionincluding aluminum content, mechanical properties (such as yield and ultimate tensilestrengths) and microstructure. The aluminum bronze samples exposed to ECW systemraw water environment will come from a pump shaft line casing pipe and from two smallcast valve bodies. The pump shaft line casing pipe was removed from service in 2012and the two small cast valve bodies will be removed from service in 2012. Thecomponents to be sampled have been exposed to ECW system raw water environmentsince the ECW system entered service. Priority will be given to selecting 100%dealloyed component samples. STP will complete this testing prior to the end of 2012.Beginning 10 years prior to the period of extended operation for each 10-year interval,periodic metallurgical testing will be performed to confirm that the load carrying capacityof aged dealloyed aluminum bronze material in the ECW system remains adequate tosupport the intended function of the system during the period of extended operation. Foreach 10-year interval beginning 10 years prior to the period of extended operation, 20percent of leaking above ground components removed from service, but at least one, willbe tested every five years. Tensile test samples from a removed component will betested to include both leaking and non-leaking portions of the component. If at least twoleaking components are not identified two years prior to the end of each 10-year testing NOC-AE-13003041Page 20 of 30interval, a risk-ranked approach based on those components most susceptible todegradation will be used to identify candidate components for removal and testing so atleast two components are tested during the 10-year interval. The component will besectioned to size the inside surface flaws, if present, and/or mapping of the dealloyedsurface areas for determining the degree of the dealloying. The samples will be testedfor chemical composition including aluminum content, mechanical properties (such asyield and ultimate tensile strengths) and microstructure. Ultimate tensile strength will betrended and compared to the acceptance criterion. The degree of dealloying andcracking will be trended by comparing examination results with previous examinationresults.An engineering evaluation will be performed at the end of each test to determine if thesample size requires adjustment based on the results of the tests. The structuralintegrity analyses will be updated as required to validate adequate load carryingcapacity.Plant procedure directs that every six months (not to exceed nine months) an inspectionof all susceptible aluminum bronze (copper alloy with greater than eight percentaluminum) above ground components be completed to identify any components thatshow evidence of dealloying. Aluminum bronze (copper alloy with greater than 8percent aluminum) components which are found to have indications of through-walldealloying are evaluated, and scheduled for replacement by the corrective actionprogram. Components g.rater than one inch will be replaced by the end of thesubsequent refueling outage.STP has buried copper alloy piping with less than eight percent aluminum that is notsusceptible to dealloying. However, there are welds in which the filler metal is copperalloy with greater than eight percent aluminum material. Therefore, the proceduredirects that a yard walkdown be performed above the buried piping aluminum bronzewelds, from the intake structure to the unit and from the unit to the discharge structure tolook for changes in ground conditions that indicate leakage. If leaking below-gradewelds are discovered by surface water monitoring or during a buried ECW pipinginspection, a section of each leaking weld will be removed for destructive examination.Preventive Actions (Element 2)The Selective Leaching of Aluminum Bronze program does not prevent degradation dueto aging effects but provides for inspections to detect aging degradation prior to the lossof intended functions, replacement of degraded components, and testing to confirm loadcarrying capacity of aged dealloyed aluminum bronze material.The Open-Cycle Cooling Water program (B2.1.9) uses an oxidizing biocide treatment(sodium hypochlorite and sodium bromide) to reduce the potential for microbiologicallyinfluenced corrosion.Parameters Monitored or Inspected (Element 3)The Selective Leaching of Aluminum Bronze program includes visual inspections everysix months (not to exceed nine months) for dealloying in all susceptible aluminum bronze(copper alloy with greater than eight percent aluminum) components. During these NOC-AE-13003041Page 21 of 30inspections, if evidence of through-wall dealloying is discovered, the components areevaluated and scheduled for replacement by the corrective action program.Components, greater than one inch, will be replaced by the end of the next refuelingoutage.During the walkdown of the buried essential cooling water piping, the ground is observedfor conditions that would indicate leakage due to selective leaching. Wheneveraluminum bronze materials are exposed during inspection of the buried essential coolingwater piping, the components are examined for indications of selective leaching. Ifleaking below-grade welds are discovered by surface water monitoring or during aburied ECW piping inspection, a section of each leaking weld will be removed fordestructive examination.Detection of Aging Effects (Element 4)The Selective Leaching of Aluminum Bronze program includes visual inspection ofaluminum bronze (copper alloy with greater than eight percent aluminum) components todetermine if selective leaching of these components is occurring. Every six months (notto exceed nine months), an inspection of susceptible above ground aluminum bronze(copper alloy with greater than eight percent aluminum) components is completed toidentify any components that show evidence of dealloying. Every 6 months, walkdown isperformed above the buried essential cooling water piping containing copper alloy weldswith an Aluminum content greater than 8 percent. During the walkdown, the soil isobserved to identify conditions that may be an indication of leakage due to selectiveleaching. Whenever aluminum bronze materials are exposed during inspection of theburied essential cooling water and ECW screen wash system piping, the componentsare examined for indications of selective leaching. If leaking below-grade welds arediscovered by surface water monitoring or during a buried ECW piping inspection, asection of each leaking weld will be removed for destructive examination.Aluminum bronze (copper alloy with greater than 8 percent aluminum) componentswhich are found to have indications of through-wall dealloying are evaluated, andscheduled for replacement by the corrective action program. Components, gieate,ene-i h-, will be replaced by the end of the next refueling outage.Volumetric examinations of aluminum bronze material components that demonstrateexternal leakage will be performed where the configuration supports this type ofexamination to conclude with reasonable assurance that cracks are not approaching acritical size.Destructive examination of each leaking component removed from service will beperformed to determine the degree of dealloying until 10 percent of the susceptiblecomponents in the ECW system are examined. The degree of dealloying and crackingwill be trended by comparing examination results with previous examination results.Metallurgical testing of leaking aluminum bronze material components in the ECWsystem removed from service will be performed to update the structural integrityanalyses, to confirm load carrying capacity and to determine the degree of dealloying bydestructive examination. Metallurgical testing of the removed leaking component will beperformed until at least three different size components of two samples each are tested, NOC-AE-1 3003041Page 22 of 30and at least nine total samples are tested. The metallurgical testing will include fracturetoughness testing of test samples that include a crack in the dealloyed material wheresufficient sample size supports bend testing. Additionally, the samples will be tested forchemical composition including aluminum content, mechanical properties (such as yieldand ultimate tensile strengths) and microstructure. Ultimate tensile strength will betrended and compared to the acceptance criterion. Degree of dealloying and crackingwill be trended by comparing examination results with previous examination results.As part of the testing described above, six samples from three aluminum bronzecomponents removed from service in 2012 will be tested for chemical compositionincluding aluminum content, mechanical properties (such as yield and ultimate tensilestrengths) and microstructure. The aluminum bronze samples exposed to ECW systemraw water environment will come from a pump shaft line casing pipe and from two smallcast valve bodies. The pump shaft line casing pipe was removed from service in 2012and the two small cast valve bodies will be removed from service in 2012. The samplecomponents have been exposed to ECW system raw water environment since the ECWsystem entered service. Priority will be given to selecting 100% dealloyed componentsamples. STP will complete this testing prior to the end of 2012.Beginning 10 years prior to the period of extended operation for each 10-year interval,periodic metallurgical testing will be performed to confirm that the load carrying capacityof aged dealloyed aluminum bronze material in the ECW system remains adequate tosupport the intended function of the system during the period of extended operation.For each 10 year interval beginning 10 years prior to the period of extended operation,20 percent of leaking above ground components removed from service, but at least one,will be tested every five years. Tensile test samples from a removed component will betested to include both leaking and non-leaking portions of the component. If at least twoleaking components are not identified two years prior to the end of each 10-year testinginterval, a risk-ranked approach based on those components most susceptible todegradation will be used to identify candidate components for removal and testing so atleast two components are tested during the 10-year interval. The component will besectioned to size the inside surface flaws, if present, and/or to map the dealloyedsurface areas for determining the degree of the dealloying. The samples will be testedfor chemical composition including aluminum content, mechanical properties (such asyield and ultimate tensile strengths) and microstructure. Ultimate tensile strength will betrended and compared to the acceptance criterion. The degree of dealloying andcracking will be trended by comparing examination results with previous examinationresults.An engineering evaluation will be performed at the end of each test to determine if thesample size requires adjustment based on the results of the tests. The structuralintegrity analyses will be updated as required to validate adequate load carryingcapacity.Monitoring and Trending (Element 5)The degree of dealloying and cracking will be trended by comparing examination resultswith previous examination results. NOC-AE-13003041Page 23 of 30The ultimate tensile strength results from the metallurgical aluminum bronze materialtesting will be monitored and trended. Trending provides monitoring of the degree ofdealloying, the degree of cracking, and the ultimate tensile strength for aging aluminumbronze material through the period of extended operation. Upon completion of eachtest, the data trended will be evaluated against the acceptance criteria for ultimatetensile strength.Acceptance Criteria (Element 6)Dealloying of aluminum bronze components is a well known phenomenon at STP. Along term improvement plan was developed in May 1992. As a result of these analyses,aluminum bronze (copper alloys with greater than eight percent aluminum) componentsare visually inspected every six months (not to exceed nine months). Upon discovery ofvisual evidence of through-wall dealloying, components are evaluated, and scheduledfor replacement by the corrective action program. Components, greater than one inch,will be replaced by the end of the next refueling outage. Due to the slow nature ofdealloying, this replacement interval provides reasonable assurance that the systemsand components within the scope of this program will continue to perform their intendedfunctions consistent with the current licensing basis for the period of extended operation.The acceptance criterion for ultimate tensile strength value of aluminum bronze materialis greater than or equal to 30 ksi. The acceptance criterion for the fracture toughness isgreater than or equal to 65 ksi in112 for aluminum bronze castings and at welded joints inthe heat affected zones. The acceptance criterion for yield strength is equal to orgreater than one-half of the ultimate strength. If a criterion is not met, the condition willbe documented in the corrective action program to perform a structural integrity analysisto confirm that the load carrying capacity of the tested material remains adequate tosupport the intended function of the ECW System through the period of extendedoperation.Corrective Actions (Element 7)STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and areacceptable in addressing corrective actions. The QA program includes elements ofcorrective action, and is applicable to the safety-related and nonsafety-related systems,structures and components that are subject to aging management review.Confirmation Process (Element 8)STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and areacceptable in addressing confirmation processes and administrative controls. TheQA program includes elements of corrective action, and is applicable to thesafety-related and nonsafety-related systems, structures and components that aresubject to aging management review.Administrative Controls (Element 9)See Element 8. NOC-AE-13003041Page 24 of 30Operating Experience (Element 10)A review of the STP plant-specific operating experience indicates that macrofouling,general corrosion, erosion-corrosion, and through-wall dealloying have been observed inaluminum bronze components. STP has analyzed the effects of the through-walldealloying and found that the degradation is slow so that rapid or catastrophic failure isnot a consideration. STP has determined that the leakage can be detected before theflaw reaches a limiting size that would affect the intended functions of the essentialcooling water and essential cooling water screen wash system. A long rangeimprovement plan and engineering evaluation were developed to deal with thedealloying of aluminum bronze components when dealloying has been identified. Basedon these analyses, the approach has been to evaluate components, and schedulereplacement by the corrective action program. Components with indications of throughwall dealloying, associated with groater thaRnG, inch Oi diameter, will be replacedby the end of the next refueling outage. A monitoring and inspection program providesconfidence in the ability to detect the leakage.EnhancementsPrior to the period of extended operation, the following enhancements will beimplemented in the following program elements:Scope of Program (Element 1)Procedures will be enhanced to:Perform volumetric examinations of aluminum bronze material components thatdemonstrate external leakage where the configuration supports this type ofexamination to conclude with reasonable assurance that cracks are not approachinga critical size.Perform destructive examination of each leaking component removed from serviceto determine the degree of dealloying until 10 percent of the susceptible componentsin the ECW system are examined. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.Prior to the period of extended operation, metallurgical testing of leaking aluminumbronze material components in the ECW system removed from service will beperformed to update the structural integrity analyses, to confirm load carryingcapacity and to determine the degree of dealloying by destructive examination.Metallurgical testing of the removed leaking component will be performed until atleast three different size components of two samples each are tested, and at leastnine total samples are tested. The metallurgical testing will include fracturetoughness testing of test samples that include a crack in the dealloyed materialwhere sufficient sample size supports bend testing. Additionally, the samples will betested for chemical composition including aluminum content, mechanical properties(such as yield and ultimate tensile strengths) and microstructure. Ultimate tensilestrength will be trended and compared to the acceptance criterion. The degree ofdealloying and cracking will be trended by comparing examination results withprevious examination results. NOC-AE-13003041Page 25 of 30As part of the testing described above, test six samples from three aluminum bronzecomponents removed from service in 2012 for chemical composition includingaluminum content, mechanical properties (such as yield and ultimate tensilestrengths) and microstructure. The aluminum bronze test samples exposed to ECWsystem raw water environment are to come from a pump shaft line casing pipe andfrom two small cast valve bodies. The pump shaft line casing pipe was removedfrom service in 2012 and the two small cast valve bodies will be removed fromservice in 2012. Priority shall be given to selecting 100% dealloyed componentsamples.Beginning 10 years prior to the period of extended operation for each 10-yearinterval, periodically test samples of above ground ECW system componentsremoved from service for chemical composition including aluminum content,mechanical properties (such as yield and ultimate tensile strengths) andmicrostructure. For each 10 year interval beginning 10 years prior to the period ofextended operation, 20 percent of leaking components removed from service, but atleast one, will be tested every five years. Tensile test samples from a removedcomponent shall be tested to include both leaking and non-leaking portions of thecomponent. If at least two leaking components are not identified two years prior tothe end of each 10-year testing interval, a risk-ranked approach will be used basedon those components most susceptible to degradation to identify candidatecomponents for removal and testing so at least two components are tested duringthe 10-year interval. The component will be sectioned to size the inside surfaceflaws, if present, and/or mapping of the dealloyed surface areas for determining thedegree of the dealloying. The samples will be tested for chemical compositionincluding aluminum content, mechanical properties (such as yield and ultimatetensile strengths) and microstructure. Ultimate tensile strength will be trended andcompared to the acceptance criterion. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.Perform an engineering evaluation at the end of each test to determine if the samplesize requires adjustment based on the results of the tests.Perform a structural integrity analysis to confirm that the load carrying capacity ofthe tested material remains adequate to support the intended function of the ECWsystem through the period of extended operation.Parameters Monitored and Inspected (Element 3)Procedures will be enhanced to indicate that whenever aluminum bronze materials areexposed during inspection of the buried essential cooling water piping, the componentsare examined for indications of selective leaching. If leaking below-grade welds arediscovered by surface water monitoring or during a buried ECW piping inspection, asection of each leaking weld will be removed for destructive metallurgical examination.Detection of Aging Effects (Element 4) NOC-AE-13003041Page 26 of 30Procedures will be enhanced to:Indicate that whenever aluminum bronze materials are exposed during inspection ofthe buried essential cooling water piping, the components are examined forindications of selective leaching. If leaking below-grade welds are discovered bysurface water monitoring or during a buried ECW piping inspection, a section ofeach leaking weld will be removed for destructive examination.Perform volumetric examinations of aluminum bronze material components thatdemonstrate external leakage where the configuration supports this type ofexamination to conclude with reasonable assurance that cracks are not approachinga critical size.Perform destructive examination of each leaking component removed from serviceto determine the degree of dealloying until 10 percent of the susceptible componentsin the ECW system are examined. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.Metallurgical testing of leaking aluminum bronze material components in the ECWsystem removed from service will be performed to update the structural integrityanalyses, to confirm load carrying capacity and to determine the degree ofdealloying by destructive examination. Metallurgical testing of the removed leakingcomponent will be performed until at least three different size components of twosamples each are tested, and at least nine total samples are tested. Themetallurgical testing will include fracture toughness testing of test samples thatinclude a crack in the dealloyed material where sufficient sample size supports bendtesting. Additionally, the samples will be tested for chemical composition includingaluminum content, mechanical properties (such as yield and ultimate tensilestrengths) and microstructure. Ultimate tensile strength will be trended andcompared to the acceptance criterion. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.As part of the testing described above, test six samples from three aluminum bronzecomponents removed from service in 2012 for chemical composition includingaluminum content, mechanical properties (such as yield and ultimate tensilestrengths) and microstructure. The aluminum bronze test samples exposed to ECWsystem raw water environment are to come from a pump shaft line casing pipe andfrom two small cast valve bodies. The pump shaft line casing pipe was removedfrom service in 2012 and the two small cast valve bodies will be removed fromservice in 2012. Priority shall be given to selecting 100% dealloyed componentsamples.Beginning 10 years prior to the period of extended operation for each 10-yearinterval, periodically test samples of above ground ECW system componentsremoved from service for chemical composition including aluminum content,mechanical properties (such as yield and ultimate tensile strengths) andmicrostructure. For each 10 year interval beginning 10 years prior to the period ofextended operation, 20 percent of leaking components removed from service, but at NOC-AE-13003041Page 27 of 30least one, will be tested every five years. Tensile test samples from a removedcomponent shall be tested to include both leaking and non-leaking portions of thecomponent. If at least two leaking components are not identified two years prior tothe end of each 10-year testing interval, a risk-ranked approach will be used basedon those components most susceptible to degradation to identify candidatecomponents for removal and testing so at least two components are tested duringthe 10-year interval. The component will be sectioned to size the inside surfaceflaws, if present, and/or mapping of the dealloyed surface areas for determining thedegree of the dealloying. The samples will be tested for chemical compositionincluding aluminum content, mechanical properties (such as yield and ultimatetensile strengths) and microstructure. Ultimate tensile strength will be trended andcompared to the acceptance criterion. The degree of dealloying and cracking will betrended by comparing examination results with previous examination results.Perform an engineering evaluation at the end of each test to determine if the samplesize requires adjustment based on the results of the tests.Perform a structural integrity analysis to confirm that the load carrying capacity ofthe tested material remains adequate to support the intended function of the ECWsystem through the period of extended operation.Monitoring and Trending (Element 5)Procedures will be enhanced to:Trend the degree of dealloying and cracking by comparing examination results withprevious examination results.Trend ultimate tensile strength results from the metallurgical aluminum bronzematerial testing.Upon completion of each test, evaluate the data trended against the acceptancecriteria for ultimate tensile strength.Acceptance Criteria (Element 6)Procedures will be enhanced to:Specify the acceptance criterion for ultimate tensile strength value of aluminumbronze material is greater than or equal to 30 ksi.Specify the acceptance criterion for fracture toughness is 65 ksi in112 for aluminumbronze castings and at welded joints in the heat affected zones.Specify the acceptance criterion for yield strength is equal to or greater than one-halfof the ultimate strength.Initiate a corrective action document when the acceptance the criterion is not met. NOC-AE-13003041Page 28 of 30ConclusionThe continued implementation of the Selective Leaching of Aluminum Bronze programprovides reasonable assurance that aging effects will be managed such that the systemsand components within the scope of this program will continue to perform their intendedfunctions consistent with the current licensing basis for the period of extended operation. NOC-AE-1 3003041Page 29 of 30A4 LICENSE RENEWAL COMMITMENTSTable A4-1 identifies proposed actions committed to by STPNOC for STP Units 1 and 2 in its License Renewal Application. Theseand other actions are proposed regulatory commitments. This list will be revised, as necessary, in subsequent amendments toreflect changes resulting from NRC questions and STPNOC responses. STPNOC will utilize the STP commitment tracking system totrack regulatory commitments. The Condition Report (CR) number in the Implementation Schedule column of the table is forSTPNOC tracking purposes and is not part of the amended LRA.Table A4-1 License Renewal CommitmentsL " =: .. .. ... ." .. .' .. ..... ; .. .. ... ... .... ..~;': , := -Item# .Commiment...... ..............................LRA ImplementationSection Schedule27 Implement the PWR Reactor Internals program as described in LRA Section B2.1.35. B2.1.35 Within 24 monthsaftor the iccuancoof EPRI 1022863,Completed__________________________________________________CR 10-2360237 Groundwater samples will be taken at multiple locations around the site every three 132.1 .32 SeptembeF-201-2months for at least 24 consecutive months. The samples will analyze for pH, sulfates, Completedand chlorides. This sampling plan will begin no later than September 2012.1_ CR 11-20856-142 Enhance the Reactor Head Closure Studs program procedures to: B2.1.3 Starting with the* perform a remote VT-1 of stud insert #30 (Unit 2 only) concurrent with the current (Thirdvolumetric examination once every 10 years to verify no additional loss of Interval) 10-yearbearing surface area. ASME Section XI NOC-AE-13003041Page 30 of 30inspection intervalCR 12-1517043 The seal cap enclosures from Unit 2 Safety Injection System Check Valve S10010A B2.1.7 2012Refiueli~and from Unit 1 and Unit 2 Chemical Volume Control System Check Valves CV0001, (Unit 1CV0002, CV0004, and CV0005 will be permanently removed. After removal of the completediseal cap enclosures, the component bolting will be replaced or inspected forintergranular stress corrosion cracking. 2013 RefuelingOutage (Unit 2)CR 12-21155

v t e Letter Sequence Other
TAC:ME4937, (Approved, Closed) TAC:ME4936, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Meeting, Meeting, Meeting Questions 8 February 2012 22 May 2012 14 April 2011 14 April 2011 14 April 2011 24 May 2011 6 July 2011 12 July 2011 28 July 2011 15 August 2011 15 August 2011 15 August 2011 18 August 2011 21 September 2011 22 September 2011 22 September 2011 4 October 2011 11 October 2011 11 October 2011 18 October 2011 3 November 2011 6 December 2011 5 December 2011 14 December 2011 6 December 2011 4 January 2012 16 February 2012 9 February 2012 30 January 2012 28 February 2012 21 March 2012 19 April 2012 12 April 2012 25 June 2012 14 June 2012 26 July 2012 12 September 2012 3 October 2012 19 November 2012 18 December 2012 6 March 2014 16 July 2014 13 April 2015 27 May 2015 28 May 2015 11 June 2015 5 October 2015 30 November 2015 30 November 2015 22 April 2016 14 June 2016 18 January 2017 ... further results Answers 25 October 2011 31 May 2012 31 May 2012 31 May 2012 6 September 2016 5 May 2011 12 May 2011 23 June 2011 9 August 2011 23 August 2011 18 August 2011 15 September 2011 15 September 2011 10 October 2011 21 November 2011 21 November 2011 4 November 2011 21 November 2011 17 November 2011 6 December 2011 15 December 2011 15 December 2011 5 January 2012 18 January 2012 6 February 2012 16 February 2012 27 February 2012 12 March 2012 5 March 2012 28 March 2012 28 March 2012 29 March 2012 17 April 2012 26 April 2012 3 May 2012 31 May 2012 14 May 2012 31 May 2012 31 May 2012 7 June 2012 14 June 2012 5 July 2012 27 June 2012 17 July 2012 4 October 2012 21 August 2012 6 December 2012 27 February 2014 18 February 2014 20 March 2014 ... further results Results Approval, Approval Other: ML110620203, ML11140A163, ML11224A265, ML11230A003, ML11318A121, ML11318A122, ML11318A123, ML11318A124, ML11318A125, ML11318A126, ML11318A127, ML11318A128, ML11318A129, ML11318A130, ML11318A131, ML11318A132, ML11318A133, ML11318A135, ML11318A136, ML11318A138, ML11318A139, ML11318A140, ML11318A141, ML11318A142, ML11318A143, ML11318A144, ML11318A145, ML11318A146, ML11318A147, ML11318A148, ML11318A149, ML11318A150, ML11318A151, ML11318A152, ML11318A153, ML11318A154, ML11318A155, ML11318A156, ML11318A157, ML11318A158, ML11318A159, ML11318A160, ML11318A161, ML11318A162, ML11318A163, ML11318A164, ML11318A165, ML11318A166, ML11318A167, ML11318A168... further results
MONTHYEAR2011-06-16 [Table View]

ML13310A311

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