ML11305A076

South Texas Project, Units 1 & 2, Response to Requests for Additional Information for License Renewal Application (TAC ME4936 and ME4937)

ML11305A076
Person / Time
Site:	South Texas
Issue date:	10/25/2011
From:	Rencurrel D W South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME4936, TAC ME4937
Download: ML11305A076 (18)

Text

Nuclear Operating CompanySouth Texas Prolect Electric Generatig Station PO Box 289 Wadsworth, Texas 77483 -October 25, 2011NOC-AE-1 100274410CFR54STI: 32991978File: G25U. S. Nuclear Regulatory CommissionAttention: Document Control DeskOne White Flint North11555 Rockville PikeRockville, MD 20852-2746South Texas ProjectUnits 1 and 2Docket Nos. STN 50-498, STN 50-499Response to Requests for Additional Information for theSouth Texas Proiect License Renewal Anolication (TAC Nos. ME4936 and References: 1. STPNOC Letter dated October 25, 2010, from G. T. Powell to NRC DocumentControl Desk, "License Renewal Application" (NOC-AE-10002607)(ML1 03010257)2. NRC letter dated September 22, 2011, "Requests for Additional Information forthe Review of the South Texas Project, Units 1 and 2 License RenewalApplication -Aging Management Review, Set 3 (TAC Nos. ME4936 andME4937)" (ML1 1258A161)By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a License RenewalApplication (LRA) for South Texas Project (STP) Units 1 and 2. By Reference 2, the NRC staffrequests additional information for review of the STP LRA. STPNOC's response to the requestfor additional information is provided in Enclosure 1 to this letter. Changes to the LRA describedin Enclosure 1 are depicted in line-in/line-out pages provided in Enclosure 2.There are no regulatory commitments in this letter.Should you have any questions regarding this letter, please contact either Arden Aldridge, STPLicense Renewal Project Lead, at (361) 972-8243 or Ken Taplett, STP License Renewal Projectregulatory point-of-contact, at (361) 972-8416.I declare under penalty of perjury that the foregoing is true and correct.Executed on /bt_____,,DateO.W. RencurrelSenior Vice President,Technical Support & OversightKJTEnclosures: 1. STPNOC Response to Requests for Additional Information2. STPNOC LRA Changes with Line-in/Line-out Annotations NOC-AE-l 1002744Page 2cc:(paper copy without enclosures)Regional Administrator, Region IVU. S. Nuclear Regulatory Commission612 East Lamar Blvd, Suite 400Arlington, Texas 76011-4125Balwant K. SingalSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North (MS 8B1)11555 Rockville PikeRockville, MD 20852Senior Resident InspectorU. S. Nuclear Regulatory CommissionP. 0. Box 289, Mail Code: MN1 16Wadsworth, TX 77483C. M. CanadyCity 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 O11F01)Washington, DC 20555-0001(electronic copy without enclosures)A. 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. RatliffAlice RogersTexas Department of State Health ServicesBalwant K. SingalJohn W. DailyTam TranU. S. Nuclear Regulatory Commission

Enclosure

1NOC-AE-1 1002744Enclosure ISTPNOC Response to Requests for Additional Information

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1NOC-AE-1 1002744Page 1 of 7SOUTH TEXAS PROJECT, UNITS I AND 2REQUESTS FOR ADDITIONAL INFORMATION -AGING MANAGEMENT REVIEW SET 3(TAC NOS. ME4936 AND ME4937)Selective Leaching (034)RAI 3.3.2.3.7-1Backgqround:In license renewal application (LRA) Table 3.3.2-7, the applicant includes an aging managementreview (AMR) item for copper alloy greater than 15 percent zinc solenoid valve internallyexposed to plant indoor air. The applicant stated that the component will be managed for loss ofmaterial using the Selective Leaching of Materials program. The AMR item lists generic note Gand plant-specific note 3, indicating that this material exposed to plant indoor air is subject towetting due to condensation, and thus is subject to loss of material due to selective leaching.The Generic Aging Lessons Learned (GALL) Report and the Metals Handbook Desk Edition(Second Edition, ASM International, 1998) both state that copper alloy greater than 15 percentzinc may be subject to stress corrosion cracking in solutions containing ammonia orammonia-like compounds such as amines provided sufficient tensile stresses are present.Issue:The staff does not have sufficient information regarding the control or use of ammonia or amines(e.g., cleaning solutions, chemicals, decay of insects) in the vicinity of the instrument air system(air intake) to determine if stress corrosion cracking should be an aging effect requiring agingmanagement.Request:Describe what measures are taken to. prevent or limit the presence of ammonia and amines inthe instrument air system.STPNOC Response:The instrument air intakes are located inside the Turbine Generator Building (TGB). The TGBcontains the instrument air compressor intake area which is a designated housekeeping areawith no standing water or nearby enclosures in which insects could build-up. Noammonia-based chemicals are used in the instrument air compressor intake area. Somecleaning solutions may contain a very low concentration of ammonia but the cleaning solutionswould be used at such a low percentage of time that it is considered insignificant. Therefore,instrument air intake piping is not exposed to airborne amines or ammonia-based compounds insufficient quantities to be of significance. A review of South Texas Project (STP) operatingexperience did not find any evidence of stress corrosion cracking associated with copper alloygreater than 15 percent zinc.

Enclosure

1NOC-AE-1 1002744Page 2 of 7Buried Piping (035)RAI 3.3.2.3.4-1Backgiround:In LRA Table 3.3.2-4, the applicant includes an AMR item for copper alloy piping (greater than 8percent aluminum) externally exposed to soil. The applicant stated that the component will bemanaged for loss of material using the Buried Piping and Tanks Inspection program. The AMRitem lists generic note G, indicating that the environment is not in the GALL Report for thematerial and environment. The Metals Handbook states that copper alloy (greater than 8 percentaluminum) may be subject to stress corrosion cracking in solutions containing ammonia orammonia-like compounds such as amines, provided sufficient tensile stresses are present.Issue:Table 2.2-1 of the Updated Final Safety Analysis Report states a chemical plant that previouslyused anhydrous ammonia is located 4.8 miles NNE of the South Texas Project (STP) site. TheEnvironmental Report (ER) states that some of the STP site east of the main cooling reservoir isleased for cattle grazing. The ER also describes the land surrounding the plant as fairly flat andused for ranchland and farmland.The plant procedure for implementing the Buried Piping and Tanks Inspection Program statesthat soil analysis data (i.e., pH, resistivity, redox potential, sulfide and sulfate ion concentration,chloride concentration, conductivity, and moisture content) should be collected duringexcavations to help assess the likelihood of pipe outside diameter corrosion. However, theprocedure does not indicate if the soil analysis tests for the presence of ammonia orammonia-like compounds.The staff does not have sufficient information regarding the presence or absence of ammonia orammonia-like compounds in the soil in and around the buried copper alloy (greater than 8percent aluminum) piping such that stress corrosion cracking could be ruled out as a possibleaging effect requiring aging management.Request:Describe what, if any, measures are taken to detect the presence or absence of ammonia in thesoil near the buried piping of interest. If it is determined that there is a potential for ammonia orammonia-like compounds to be present in the soil in the vicinity of the piping of interest, describewhat measures will be taken to manage stress corrosion cracking of the buried copper alloy(greater than 8 percent aluminum) piping.STPNOC Response:The outdoor environment at STP is not subject to industrial pollution. The closest industrialfacility is 4.8 miles away. There have been no industrial ammonia events detected at the STPplant site. The land adjacent to the plant site is open range for cattle and farming. There are nolarge concentrations of cattle within five miles of the plant site that could generate excessivedetrimental gases or concentrated solid waste. There is no runoff from adjacent land onto theplant site. A search of the STP corrective action database did not identify any ammonia orammonia-like compound spills or contaminations that affected on-site soil conditions. There isno evidence to expect the soil at STP to have elevated levels of ammonia or ammonia-like

Enclosure

1NOC-AE-1 1002744Page 3 of 7compounds. Therefore, it is not necessary to consider ammonia-related aging effects on buriedcopper alloy (>8% aluminum) components.Diesel Exhaust Piping (078)RAI 3.3.2.2.3.3-1Background:LRA Table 3.3.2-21 includes stainless steel expansion joints exposed to diesel exhaust (internal)for the nonsafety-related diesel generator that are being managed for loss of material. For thecorresponding material and environment, the GALL Report recommends managing for both lossof material and cracking due to stress corrosion cracking, and recommends using aplant-specific AMP.Issue:The stainless steel expansion joint exposed to diesel exhaust in LRA Table 3.3.2-21 is not beingmanaged for stress corrosion cracking as recommended by the GALL Report.Request:Provide the basis for not managing the stainless steel expansion joint exposed to diesel exhaustin Table 3.3.2-21 for stress corrosion cracking or provide a suitable AMP that will manage thisaging effect for this material and environment combination.STPNOC Response:LRA Table 3.3.2-21 and Section 3.3.2.1.21 will be revised to add an aging effect of cracking forstainless steel expansion joints exposed to an internal environment of diesel exhaust usingNUREG 1801 AMR line VII.H2-1. Aging management program B2.1.22, Inspection of InternalSurfaces in Miscellaneous Piping and Ducting Components, will be used to manage the crackingof stainless steel expansion joints exposed to an internal environment of diesel.Enclosure 2 provides the line-in/line-out sections of the License Renewal Application.Boric Acid Corrosion (010)RAI 3.1.1.58-1Background:In LRA Tables 3.1.2-1, 3.1.2-2, 3.1.2-3, and 3.1.2-4, the applicant stated that several steelcomponent external surfaces exposed to borated water leakage are managed for loss ofmaterial by the Boric Acid Corrosion Program (LRA Section B2.1.4). These items are associatedwith LRA Table 3.1-1, item 3.1.1.58.The updated staff guidance in SRP-LR, Revision 2, Table 3.1-1, item 48, states that steelexternal surfaces, including reactor vessel top head, bottom head, and reactor coolant pressureboundary piping or components adjacent to dissimilar metal welds exposed to air with boratedwater leakage, should be managed for loss of material due to boric acid corrosion by GALL

Enclosure

1NOC-AE-1 1002744Page 4 of 7AMPs XI.M10, "Boric Acid Corrosion" and XI.M1 1 B, "Cracking of Nickel-Alloy Components andLoss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure BoundaryComponents."The GALL AMP XI.M1 1 B "scope of program" program element states that this programmanages loss of material due to boric acid corrosion in steel components in the vicinity ofnickel-alloy components, including, but not limited to, reactor vessel components, steamgenerator components, pressurizer components, and reactor coolant system piping. Theprogram description states that inspection activities should be in accordance with10 CFR 50.55a, including ASME Code Cases N-722-1 and N-729-1, and industry guidelines forinspection of primary system butt welds (e.g. MRP-1 39).Issue:The program description in LRA Section B2.1.4, "Boric Acid Corrosion," refers to inserviceinspections in accordance to ASME Code Section Xl; however, it is not clear to the staff whetherthe requirements in 10 CFR 50.55a, including Code Cases N-722-1 and N-729-1, and MRP-1 39,are incorporated in those inspections.Request:Clarify whether the inservice inspections in the Boric Acid Corrosion Program are in accordancewith 10 CFR 50.55a, including ASME Code Cases N-722-1 and N-72901, and MRP-139. If not,provide information on what equivalent inspection activities will be used to manage loss ofmaterial due to boric acid corrosion of steel components in the vicinity of nickel-alloy reactorcoolant pressure boundary components.STPNOC Response:The inspection activities identified by 10 CFR 50.55a, including ASME Code Cases N-722-1 andN-729-1, and MRP-1 39, pertain to the detection of cracking of nickel-alloy components(including welds) and do not pertain to the loss of material of carbon steel components. At STPthe aging effect of loss of material due to boric acid corrosion for steel components in the vicinityof nickel-alloy components is managed by the Boric Acid Corrosion program (B2.1.4). The BoricAcid Corrosion program includes provisions to identify leakage, inspect and examine forevidence of leakage, evaluate the effects of leakage, and initiate corrective actions, asappropriate.The aging effect of cracking of susceptible nickel-alloy components (including welds) of thereactor pressure boundary is managed by Penetration Nozzles Welded to the Upper ReactorVessel Closure Heads of Pressurized Water Reactors program (B2.1.5), and plant-specificprogram Nickel Alloy Aging Management program (B2.1.34). Aging management programB2.1.5 requires implementation of ASME Code Case N-729-1, subject to the conditions specifiedin 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6), in the ASME Section Xl, In-Service Inspections.Aging management program B2.1.34 requires implementation of ASME code case N-722,subject to the conditions listed in 10 CFR 50.55a(g)(6)(ii)(E)(2) through (4), into the ASMESection Xl, In-Service Inspections. Aging management program B2.1.34 also implementsexaminations consistent with MRP-1 39.As discussed above, ASME code cases N-729-1 and N-722 (subject to the conditions of1OCFR50.55a) and MRP-139 are applied to management of aging of Nickel Alloy components,similar to GALL Rev 2 XI.M1 1 B requirements.

Enclosure

1NOC-AE-1 1002744Page 5 of 7RAI 3.3.1.88-1Background:SRP-LR, Revision 1, Table 3.3-1, item 88 states that aluminum and copper alloy greater than 15percent Zn piping, piping components, and piping elements exposed to air with borated waterleakage should be managed for loss of material due to borated water leakage by GALL AMPXI.M10, "Boric Acid Corrosion." LRA Table 3.3.1, item 3.3.1.88 states that this item is notapplicable because there is no in-scope aluminum or copper alloy greater than 15 percent Znpiping, piping components, or piping elements exposed to air with borated water leakage in theauxiliary systems.LRA Section 3.3.2.1.19 states that the chemical and volume control system (CVCS), an auxiliarysystem, contains an environment of borated water leakage. The staff noted that in LRA Table3.3.2-19, the AMR results for the CVCS include an item for aluminum insulation; however, theonly environment cited is plant indoor air (external).Issue:Given that borated water leakage is a recognized environment in the CVCS, it is not clear to thestaff why the aluminum insulation in this system is not managed for loss of material due to boricacid corrosion.Request:Clarify whether the aluminum insulation in the chemical and volume control system may beexposed to borated water leakage. If so, state how loss of material due to boric acid corrosionwill be managed.STPNOC Response:Aging management evaluation for a treated borated water leakage environment is consideredapplicable only for components that contain treated borated water, and is not applicable foradjacent system components or insulation on the piping that contains the treated borated water.It is possible that the aluminum sheathing could be exposed to treated borated water leakage.The loss of material due to boric acid corrosion caused by treated borated water leakage fromthe system to the aluminum sheathing is managed by the Boric Acid Corrosion program asdescribed in LRA Appendix B2.1.4. The Boric Acid Corrosion program covers mechanical,electrical, and structural components made of materials susceptible to boric acid corrosion onwhich borated water leakage may occur and evaluates any components with evidence of boricacid exposure, including insulation aluminum sheathing.Reactor Head Closure Studs Program (003)RAI B2.1.3-4Background:SRP-LR, Revision 2, Table 3.0-1 addresses aging management programs used to manage theaging effects associated with various systems and the descriptions of the programs, which areacceptable for the UFSAR supplement. Specifically, SRP-LR, Revision 2, Table 3.0-1 addresses

Enclosure

1NOC-AE-1 1002744Page 6 of 7the UFSAR supplement description of GALL AMP XI.M3, "Reactor Head Closure Studs," byreferring to the inservice inspections in conformance with the requirements of the ASME Code,Section Xl, Subsection IWB, Table IWB-2500-1 and preventive measures to mitigate cracking.SRP-LR, Revision 2, Table 3.0-1 further states that the program also relies on recommendationsto address reactor head stud bolting degradation as delineated in NUREG-1339 and NRCRegulatory Guide (RG) 1.65. NUREG-1339 and RG 1.65 indicate that molybdenum sulfide is apotential contributor to stress corrosion cracking (SCC). NUREG-1 339 and RG 1.65 (Revision 1,April 2010) also include guidance for the yield strength levels of the bolting material resistant toSCC.In comparison, LRA Section A1.3 provides the UFSAR supplement description for LRA SectionB2.1.3, "Reactor Head Closure Studs Program." This LRA Section states that the applicant'sprogram follows the preventive measures in RG 1.65. However, the UFSAR supplementdescribed in LRA Section A1.3 does not include the statement that the applicant's program relieson recommendations to address reactor head stud bolting degradation as delineated inNUREG-1339 and NRC RG 1.65.Issue:In contrast with SRP-LR, Revision 2, Table 3.0-1, the applicant's UFSAR supplement for theReactor Head Closure Studs Program (described in LRA Section A1.3) does not include thestatement that the applicant's program relies on recommendations to address reactor head studbolting degradation as delineated in NUREG-1339 and NRC RG 1.65. The licensing basis forthis program for the period of extended operation may not be adequate if the applicant does notincorporate this information in its UFSAR supplement.Request:Revise the applicant's UFSAR supplement description for the Reactor Head Closure StudsProgram to be consistent with the UFSAR supplement described in SRP-LR, Revision 2, Table3.0-1, which incorporates recommendations in NUREG-1339 and NRC RG 1.65.If the applicant has determined that a revision to the UFSAR supplement description is notnecessary, justify why the omission of the information from the UFSAR supplement, regardingNUREG-1339 and NRC RG 1.65, is acceptable to provide an adequate licensing basis for thisprogram for the period of extended operation.STPNOC Response:The Reactor Head Closure Studs program manages cracking and loss of material by conductingASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closurestuds, nuts, washers, and bushings. The program includes periodic visual, surface, andvolumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs,nuts, washers, and bushings and the performance of visual inspections of the reactor vesselflange closure during primary system leakage tests. The program implements recommendationsin NUREG-1 339 and NRC Regulatory Guide 1.65 to address reactor head stud boltingdegradation except for yield strength of existing bolting materials. The program implementsASME Section X1 Code, Subsection IWB, and detects reactor vessel stud, nut, washer, andbushing cracking, loss of material due to wear and corrosion, and reactor coolant leakage fromthe reactor pressure vessel flange. STP will use the ASME Code edition consistent with theprovisions of 10 CFR 50.55a during the period of extended operation.

Enclosure

1NOC-AE-1 1002744Page 7 of 7LRA Appendices A1.3 and B2.1.3 will be revised to include reference to NUREG-1 339 and NRCRegulatory Guide 1.65.Enclosure 2 provides the line-in/line-out sections of the License Renewal Application.

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2NOC-AE-1 1002744Enclosure 2STPNOC LRA Changes with Line-inlLine-out Annotations

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2NOC-AE-1 1002744Page 1 of 7List of Revised LRA Sections-Affected LRA Section RAISection 3.3.2.1.21 3.3.2.2.3.3-1Table 3.3.2-21 3.3.2.2.3.3-1Appendix A1.3 B2.1.3-4Appendix B2.1.3 B2.1.3-4South Texas ProjectLicense Renewal ApplicationAmendment 6

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2NOC-AE-1 1002744Page 2 of 73.3.2.1.21Nonsafety-related Diesel Generators and Auxiliary Fuel Oil SystemAging Effects Requiring ManagementThe following nonsafety-related diesel generators and auxiliary fuel oil system aging effectsrequire management:* Crackinq* Loss of materialLoss of preloadSouth Texas ProjectLicense Renewal ApplicationAmendment 6

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2NOC-AE-1 1002744Page 3 of 7Table 3.3.2-21Auxiliary Systems -Summary of Aging Management Evaluation -Nonsafety-related Diesel Generators andAl i tvili;n~ t:"m n1 Qi/olfnm__ _ __ _ _ __ !. ~ '.~~L ~~" _ _ __ _ _ __ _ _ __ _ __ _ _ __ _ __ _ _ ______ __ _ _Component Type 1Intended Material Environment Aging Effect Aging Management, NUREG- Table.1 NotesFunction Requiring Program f1801 Vol. Item____ __ ___ ___ ___ ___ Management __ _ __ _ Item __ _ __Closure BoltingExpansion JointPBtp._BBCarbonSteelStainlessSteelPlant Indoor Air(.Ext)Diesel ExhaustDiesel Exhaust(Int)LiCExpansion Joint PBoss of preload 'Bolting Integrity1(B2.1.7)_racking .Inspection of Internal \Surfaces inMiscellaneous Piping,and Ducting_ Components (B2.1.22)oss of material Inspection of Internal \'Surfaces in'Miscellaneous Piping.and Ducting'Components (B2.1.22)/11.1-5/Il. /11.H2-:3.3.1.45StainlessSteelLcB1 3.3.1.062 3.3.1.18ESouth Texas ProjectLicense Renewal ApplicationAmendment 6

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2NOC-AE-1 1002744Page 4 of 7A1.3REACTOR HEAD CLOSURE STUDSThe Reactor Head Closure Studs program manages cracking and loss of material by conductingASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closurestuds, nuts, washers, and bushings. The program includes periodic visual, surface, andvolumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs,nuts, washers, and bushings and performs visual inspections of the reactor vessel flange closureduring primary system leakage tests. The pro-gram implements recommendations inNUREG-1339 and NRC Regulatory Guide 1.65 to address reactor head stud bolting degradationexcept for yield strength of existing bolting materials. The program implements ASMESection Xl code, Subsection IWB, and detects reactor vessel stud, nut, washer, and bushingcracking, loss of material due to wear and corrosion, and reactor coolant leakage from thereactor vessel flange. STP will use the ASME Code edition consistent with the provisions of10 CFR 50.55a during the period of extended operation.South Texas ProjectLicense Renewal ApplicationAmendment 6

Enclosure

2NOC-AE-1 1002744Page 5 of 7B2.1.3 Reactor Head Closure StudsProgram DescriptionThe Reactor Head Closure Studs program manages cracking and loss of material by conductingASME Section Xl inspections of reactor vessel flange stud hole threads, reactor head closurestuds, nuts, washers, and bushings. The program includes periodic visual, surface, andvolumetric examinations of reactor vessel flange stud hole threads, reactor head closure studs,nuts, washers, and bushings and performs visual inspections of the reactor vessel flange closureduring primary system leakage tests. The STP program implements ASME Section Xl code,Subsection IWB, 2004 Edition. Reactor vessel flange stud hole threads, reactor head closurestuds, nuts, washers, and bushings are identified in ASME Section Xl Tables IWB-2500-1 andare within the scope of license renewal. ST.P follows the preventive m.easures The programimplements recommendations in NUREG-1339 and iR-NRC Regulatory Guide 1.65, Material andInspection for Reactor Vessel Closure Studs, to address reactor head stud boltinq degradationexcept for yield strength of existing bolting materials. STP uses lubricants on reactor headclosure stud threads after reactor head closure stud, nut, and washer cleaning and examinationsare complete. The lubricants are compatible with the stud material and operating environmentand do not include 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 during eachsuccessive 120-month inspection interval to comply with the requirements of the latest edition ofthe Code specified twelve months before the start of the inspection interval. STP will use theASME Code Edition consistent with the provisions of 10 CFR 50.55a during the period ofextended operation.Potential cracking and loss of material conditions in reactor vessel flange stud hole threads,reactor head closure studs, nuts, washers, and bushings are detected through visual, surface, orvolumetric examinations in accordance with ASME Section Xl requirements in STP proceduresevery ten years. These inspections are conducted during refueling outages. Reactor vesselstuds are removed from the reactor vessel flange each refueling outage. Studs, nuts, washers,and bushings are stored in protective racks after removal. Reactor vessel flange holes areplugged with water tight plugs during cavity flooding. These methods assure the holes, studs,nuts, washers, and bushings are protected from borated water during cavity flooding. Reactorvessel flange leakage is detected prior to reactor startup during reactor coolant system pressuretesting each refueling outage. The STP program has proven to be effective in preventing anddetecting potential aging effects of reactor vessel flange stud hole threads, 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.South Texas ProjectLicense Renewal ApplicationAmendment 6

Enclosure

2NOC-AE-1 1002744Page 6 of 7Exceptions to NUREG-1801Proqram Elements Affected:Scope of Program (Element 1)Regulatory Guide 1.65 states that the ultimate tensile strength of stud bolting material should notexceed 170 ksi. One closure head insert has a tensile strength of 174.5 ksi. STP creditsinservice inspections that are within the scope of this AMP, which are implemented inaccordance with the STP Inservice Inspection Program, Examination Category B-G-1requirements, as the basis for managing cracking in these components. This is in accordancewith the "parameters monitored or inspected" and "detection of aging effects" program elementsin NUREG 1801,Section XI.M3. In addition, the studs, nuts and washers are coated with alubricant which is compatible with the stud materials, and the studs, nuts, and washers areprotected from exposure to boric acid by removing them and plugging the reactor vessel flangeholes during cavity flooding.Corrective Actions (Element 7)NUREG-1801,Section XI.M3 specifies the use of Regulatory Guide 1.65 requirements forclosure stud and nut material. STP uses SA-540, Grade B-24 (as modified by Code Case 1605)stud material. The use of this material has been found acceptable to the NRC for thisapplication within the limitations discussed in Regulatory Guide 1.85, Materials Code CaseAcceptability.EnhancementsNoneOperating ExperienceReview of plant-specific operating experience has not revealed any program adequacy issueswith 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 or IGSCC have beenidentified with STP reactor vessel studs, nuts, washers, bushings, and flange stud holes.Review of the Refueling Outage Inservice Inspection Summary Reports for Interval 2 indicatesthere were no repair/replacement items identified with reactor vessel closure studs, nuts,washers, bushings, or flange thread holes. None of the repair/replacement items indicate anyimplementation issues with the STP ASME Section Xl Program for reactor closure studs, nuts,washers, bushings, or flange thread holes.The ISI Program at STP is updated to account for industry operating experience. ASMESection Xl is also revised every three years and addenda issued in the interim, which allows thecode to be updated to reflect operating experience. The requirement to update the ISI Programto reference more recent editions of ASME Section Xl at the end of each inspection intervalensures the ISI Program reflects enhancements due to operating experience that have beenincorporated into ASME Section Xl.South Texas ProjectLicense Renewal ApplicationAmendment 6

Enclosure

2NOC-AE-1 1002744Page 7 of 7ConclusionThe continued implementation of the Reactor Head Closure Studs program provides reasonableassurance that aging effects will be managed such that the systems and components within thescope of this program will continue to perform their intended functions consistent with the currentlicensing basis for the period of extended operation.South Texas ProjectLicense Renewal ApplicationAmendment 6