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{{Adams
#REDIRECT [[L-11-238, Reply to Request for Additional Information for the Review of License Renewal Application and License Renewal Application Amendment No. 13]]
| number = ML11231A966
| issue date = 08/17/2011
| title = Davis-Besse, Unit 1, Reply to Request for Additional Information for the Review of License Renewal Application and License Renewal Application Amendment No. 13
| author name = Byrd K W
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000346
| license number = NPF-003
| contact person =
| case reference number = L-11-238, TAC ME4640
| document type = Letter
| page count = 144
| project = TAC:ME4640
| stage = RAI
}}
 
=Text=
{{#Wiki_filter:FENOC Davis-Besse Nuclear Power Station5501 N. State Route 2FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449August 17, 2011L-1 1-238 10 CFR 54ATTN: Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555-0001SUBJECT:Davis-Besse Nuclear Power Station, Unit No. 1Docket No. 50-346, License Number NPF-3Reply to Request for Additional Information for the Review of the Davis-Besse NuclearPower Station, Unit No. 1, License Renewal Application (TAC No. ME4640) andLicense Renewal Application Amendment No. 13By letter dated August 27, 2010 (Agencywide Documents Access and ManagementSystem (ADAMS) Accession No. ML102450565), FirstEnergy Nuclear OperatingCompany (FENOC) submitted an application pursuant to Title 10 of the Code of FederalRegulations, Part 54 for renewal of Operating License NPF-3 for the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS). By letters dated July 12, 2011(ML11189A043), July 21, 2011 (ML11195A020), July 21, 2011 (ML11196A127), andJuly 27, 2011 (ML1 1203A080), the Nuclear Regulatory Commission (NRC) requestedadditional information to complete its review of the License Renewal Application (LRA).The content and submittal date of this letter was discussed during telephone conferenceswith Mr. Samuel Cuadrado de Jesus, NRC Project Manager, and, due to the expandedscope of the letter, the submittal date was deferred to a mutually agreeable submittal dateof August 17, 2011. Attachments 1 through 4 provide the FENOC reply to the NRCrequest for additional information (RAI) by letter, with selected RAIs listed, as follows:Attachment 1: 3 of 3 RAls in NRC letter dated July 12, 2011 (ML11189A043)Attachment 2: 7 of 13 RAls in NRC letter dated July 21, 2011 (ML1 1195A020):-Includes RAIs B.2.22-7; B.2.39-10; B.2.40-2; 3.5.2.3.12-3;B.2.25-7; B.2.25-8; and, 2.3.3.18-3Attachment 3: 4 of 4 RAls in NRC letter dated July 21, 2011 (ML11196A127)Attachment 4: 3 of 5 RAls in NRC letter dated July 27, 2011 (ML1 1203A080):-Includes RAls B.2.3-5; B.2.28-1; and, B.2.36-5Attachment 5 provides the FENOC reply to NRC RAls discussed during telephoneconference calls with the NRC, and requests from NRC Region III Inspectors during the Davis-Besse Nuclear Power Station, Unit No. 1L-1 1-238Page 2NRC Inspection Procedure (IP) 71002 License Renewal Inspection heldApril 25 -29, 2011, and May 9 -13, 2011.For all Attachments, the NRC request is shown in bold text followed by theFENOC response.The Enclosure provides Amendment No. 13 to the DBNPS LRA.There are no regulatory commitments contained in this letter. If there are any questionsor if additional information is required, please contact Mr. Clifford I. Custer, FleetLicense Renewal Project Manager, at 724-682-7139.I declare under penalty of perjury that the foregoing is true and correct. Executed onAugust 17, 2011.Sincerely,Kendall W.Attachments:1. Reply to Request for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. I (DBNPS), License Renewal Application,from NRC Letter dated July 12, 2011 (ML11189A043)2. Reply to Request for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Letter dated July 21, 2011 (ML11195A020)3. Reply to Request for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC letter dated July 21, 2011 (ML11196A127)4. Reply to Request for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Letter dated July 27, 2011 (ML11203A080)5. Reply to Request for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Conference Calls and NRC Region III License Renewal InspectionEnclosure:Amendment No. 13 to the DBNPS License Renewal Application Davis-Besse Nuclear Power Station, Unit No. 1L-1 1-238Page 3cc: NRC DLR Project ManagerNRC Region III Administratorcc: w/o Attachments or EnclosureNRC DLR DirectorNRR DORL Project ManagerNRC Resident InspectorUtility Radiological Safety Board
 
==Attachment==
1L-1 1-238Reply to Request for Additional Information for the Review of theDavis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Letter dated July 12, 2011 (ML11189A043)Page 1 of 6Question RAI 3.2.2.2.3.6-2Background:By letter dated May 2, 2011, the U.S. Nuclear Regulatory Commission (NRC or thestaff) issued Request for Additional Information (RAI) 3.2.2.2.3.6-1 requesting thatFirstEnergy Nuclear Operating Company (the applicant) provide justification forits use of the One-Time Inspection Program for managing loss of material due topitting and crevice corrosion of the internal surfaces of stainless steel piping,piping components, piping elements, and tanks exposed to condensation. In itsresponse dated June 3, 2011, the applicant stated that Amendment No. 7 to thelicense renewal application (LRA) changed the aging management program(AMP) used for the condensation environment to the Inspection of InternalSurfaces in Miscellaneous Piping and Ducting Program. The applicant alsostated that the One-Time Inspection Program will still be used to verify theeffectiveness of the AMPs credited for managing aging effects above and belowthe air/water interface.The staff noted that, for the 13 aging management review (AMR) items that werethe subject of RAI 3.2.2.2.3.6-1 (items that reference LRA Table 3.2.1,item 3.2.1-08), the applicant:" Changed the AMP to the Inspection of Internal Surfaces in MiscellaneousPiping and Ducting Program for five of the items* Retained the One-Time Inspection Program for two of the items (LRATable 3.3.2-4, item 158 and Table 3.3.2-5, item 59), citing plant-specificnote 0313, which states that the One-Time Inspection Program willconfirm the absence of aging effects, or that aging is acting slowly, at theair-water interface" Did not specifically address the remaining six itemsIssue:The staff agrees with the applicant's amended position which manages stainlesssteel components exposed to condensation for loss of material due to pittingandcrevice corrosion with the Inspection of Internal Surfaces in Miscellaneous
 
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1L-1 1-238Page 2 of 6Piping and Ducting Program. However, the staff has identified the followingissues:1. For the two items for which the One-Time Inspection Program wasretained, the applicant does not age manage the internal surfaces abovethe air/water interface (i.e., there are no AMR items for the upper portionsof the associated tanks). The staff noted that these items originallyaddressed the upper portions of the tank internal surfaces exposed tocondensation, but Amendment No. 7 changed these items to specificallyage manage the air/water interface.2. For the six items that Amendment No. 7 did not address, it is unclear to thestaff how the applicant will manage loss of material due to pitting andcrevice corrosion. These items include components in the containmentspray system, core flooding system, decay heat removal and LPI system,component cooling water system, and demineralized water storage system.Request:1. State how loss of material will be managed for the internal surfaces ofLRA Table 3.3.2-4, item 158 and Table 3.3.2-5, item 59 above theair-water interface and subject to condensation.2. State how loss of material will be managed for those items thatreference LRA Table 3.2.1, item 3.2.1-08 but were not addressed inAmendment No. 7.RESPONSE RAI 3.2.2.2.3.6-2The LRA is revised as described below to credit the Inspection of Internal Surfaces inMiscellaneous Piping and Ducting Program for management of loss of material forstainless steel in a moist air environment:1. Loss of material for the internal surfaces of LRA Table 3.3.2-4, item 158, andTable 3.3.2-5, item 59, subject to condensation is managed by the Inspection ofInternal Surfaces in Miscellaneous Piping and Ducting Program.2. Loss of material for those items that reference LRA Table 3.2.1, item 3.2.1-08 ismanaged by the Inspection of Internal Surfaces in Miscellaneous Piping andDucting Program. The Inspection of Internal Surfaces in Miscellaneous Pipingand Ducting Program is also added to the list of Aging Management Programsin LRA Sections 3.2 and 3.3 for affected systems.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
==Attachment==
1L-1 1-238Page 3 of 6Question RAI 3.3.1.49-2Background:By letter dated May 2, 2011, the staff issued RAI 3.3.1.49-1 requesting that theapplicant state why loss of material due to microbiologically influenced corrosion(MIC) is not an applicable aging effect for stainless steel heat exchangercomponents exposed to closed cycle cooling water. In its response datedJune 3, 2011, the applicant stated that, because Davis-Besse has no plant-specific operating experience of MIC in its closed cooling water environments,MIC is not an aging effect requiring management.The EPRI Closed Cooling Water Chemistry Guideline, Revision 1 (1007820) statesthat MIC is a significant issue in closed cooling water systems. EPRI 1007820 andthe EPRI Non-Class 1 Mechanical Implementation Guideline and MechanicalTools, Revision 4 (1010639) also state that stagnant loops in closed cooling watersystems can accumulate microorganisms and their nutrients, and waterchemistry in these areas may be difficult to maintain.SRP-LR Section A.1.2.1, "Applicable Aging Effects," states that an aging effectshould be identified as applicable for license renewal even if there is a preventionor mitigation program associated with that aging effect. GALL AMP XI.M21A"Closed Treated Water Systems" states that, because the control of waterchemistry may not be fully effective in mitigating aging effects, visual inspectionsare conducted.Issue:The staff noted that the lack of plant-specific operating experience associatedwith MIC may be attributable to water chemistry controls and/or the absence ofinspections to specifically identify this aging mechanism in areas most prone toMIC (e.g., stagnant loops). The staff also noted that, although water chemistrycontrols may be responsible for the absence of MIC, the aging effect is stillapplicable and inspections are appropriate to ensure that the control of waterchemistry remains fully effective.Request:Include monitoring for MIC in the Closed Cooling Water Chemistry Program toensure that the control of water chemistry remains fully effective at preventingthis aging mechanism or provide technical justification for why MIC is notcredible at Davis-Besse, regardless of water chemistry controls.
 
==Attachment==
1L-1 1-238Page 4 of 6RESPONSE RAI 3.3.1.49-2Consistent with the current EPRI water chemistry guidelines, the Closed Cooling WaterChemistry Program monitors specific parameters to assure that corrosion is minimized,microbial activity is suppressed, and corrosion inhibitor stability is maintained. Theparameters that are measured include hydrazine, chloride, fluoride, biological activity,corrosion inhibitor and sulfate concentrations, as required.To clarify the intent of the Closed Cooling Water Chemistry Program, the programdescriptions, LRA Sections A.1.8 and B.2.8, both titled "Closed Cooling WaterChemistry Program," are revised to specifically state that monitoring for microbiologicalactivity in accordance with the EPRI Closed-Cycle Cooling Water guidelines is afunction of the program.Item Number 3.3.1-49 in LRA Table 3.3.1 is revised to state that loss of materialdue to microbiologically influenced corrosion is monitored by the Closed CoolingWater Chemistry Program in accordance with the EPRI Closed-Cycle CoolingWater guidelines.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI 3.3.2.2.5-2Background:In its response to RAI 3.3.2.2.5-1, dated May 24, 2011, regarding agingmanagement of elastomeric components the applicant developed a new plant-specific Inspection of Internal Surfaces in Miscellaneous Piping and DuctingProgram; however, the applicant did not revise the External Surfaces MonitoringProgram to include physical manipulation of the external surfaces of elastomers.The amended LRA contains elastomeric components exposed to air-indooruncontrolled (external) environment in Tables 3.2.2-1 and 3.3.2-1, 6, 12, 13, 14, 15,21, and 30.The GALL Report Revision 2, recommends that elastomeric materials exposed toair-indoor uncontrolled (external) be managed for hardening and loss of strengthby GALL AMP XI.M36, "External Surfaces Monitoring of MechanicalComponents." GALL AMP XI.M36 recommends that elastomeric materials bephysically manipulated during inspections to detect hardening and loss ofstrength and that the manipulation should include 10 percent of the availablesurface area.
 
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1L-11-238Page 5 of 6The GALL Report Revision 2 recommends that elastomeric materials exposed toraw water be managed by GALL AMP XI.M20, "Open-Cycle Cooling WaterSystem" which states that elastomeric components be periodically examinedconsistent with the examinations described in GALL AMP XI.M38. The staff notedthat the applicant revised its LRA AMR line items addressing elastomericmaterials exposed to raw water to be managed for hardening and loss of strengthby a newly-developed plant-specific Inspection of Internal Surfaces inMiscellaneous Piping and Ducting Program.Issue:The staff agrees with the applicant's position that elastomeric componentsexposed to an air-indoor uncontrolled (internal) environment be managed by theInspection of Internal Surfaces in Miscellaneous Piping and Ducting Program forhardening and loss of strength. However, for elastomeric components exposed toan air-indoor uncontrolled (external environment), the GALL Report recommendsthat the components be periodically inspected using a physical manipulationmethod. In addition, the applicant should state the minimum available surfacearea that will be manipulated during inspections when utilizing the ExternalSurfaces Monitoring Program.The staff does not agree that elastomeric components exposed to raw water canbe adequately managed for hardening and loss of strength by the plant-specificInspection of Internal Surfaces in Miscellaneous Piping and Ducting Programbecause it is opportunistic and therefore may not ensure periodic inspections ofelastomeric material are conducted. The staff believes periodic inspections arenecessary due to the changing nature of raw water environments.Request:a) Revise the External Surfaces Monitoring Program to include physicalmanipulation of elastomeric materials and/or:1. State how it would be effective at determining if hardening or lossof strength has occurred in the absence of physical manipulation,or2. Propose a periodic inspection program that will physicallymanipulate elastomeric components.b) Propose a periodic inspection program which includes physicalmanipulation of elastomeric components exposed to raw water, or statewhy periodic inspections are not necessary to adequately detecthardening and loss of strength in these materials.c) State the minimum available surface area that will be manipulated duringinspections of elastomeric materials. If the minimum available surfacearea that will be manipulated during inspections of elastomeric materials
 
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1L-1 1-238Page 6 of 6is less than 10 percent, state the basis for how the inspection willsufficiently identify the hardening and loss of strength aging effects.RESPONSE RAI 3.3.2.2.5-2a) The External Surfaces Monitoring Program is revised to include physicalmanipulation of elastomeric materials.b) The Collection, Drainage and Treatment Components Inspection Program, aperiodic inspection program (see response to RAI B.2.9-3 in FENOC Letterdated July 22, 2011 (ML1 1208C274)), is revised to include physicalmanipulation of elastomeric components exposed to raw water. In addition,LRA Chapter 3 aging management review line items addressing elastomericmaterials exposed to raw water are revised to show that hardening and loss ofstrength and loss of material will be managed by the Collection, Drainage andTreatment Components Inspection Program.c) As provided in the revised External Surfaces Monitoring Program and therevised Collection, Drainage and Treatment Components Inspection Program,the minimum available surface area that will be manipulated during inspectionsof elastomeric materials is 10 percent.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
==Attachment==
2L-1 1-238Reply to Request for Additional Information for the Review of theDavis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Letter dated July 21, 2011 (ML111 95A020)Page 1 of 15Question RAI B.2.22-7Background:By letter dated May 24, 2011, the applicant responded to a staff RAI B.2.22-4regarding the examination of the containment steel penetration sleeves,dissimilar metal welds, and steel components. The applicant stated that inaccordance with Title 10 of the Code of Federal Regulations(10 CFR) 50.55a(b)(2)(ix), the examinations of the Category E-B pressure retainingwelds and Category E-F pressure retaining dissimilar metal welds are notscheduled since these examinations are optional. However, the inserviceInspection (ISI) -IWE Program does include the Category E-A examinations ofContainment surfaces. Additionally, the 10 CFR Part 50, Appendix J Programdetects evidence of leakage as part of the Category E-P examinations.Issue:Davis-Besse Nuclear Power Station (DBNPS) LRA Section 4.6.2 states a search ofthe DBNPS current licensing basis did not identify any pressurization cycles orfatigue analyses for containment penetration assemblies. GALL Report, Rev. 2,AMP, XI.S1, "ASME Section XI, Subsection IWE," recommends that stainless steelpenetration sleeves, dissimilar metal welds, bellows, and steel components thatare subject to cyclic loading but have no current licensing basis fatigue analysisare monitored for cracking.Request:Please advise how DBNPS is going to comply with the GALL Reportrecommendations concerning the inspection of the containment stainless steelpenetration sleeves, dissimilar metal welds, bellows, and steel components thatare subject to cyclic loading but have no current licensing basis fatigue analysis.RESPONSE RAI B.2.22-7The Inservice Inspection (ISI) Program -IWE is revised to include an enhancement tomonitor for cracking of Containment stainless steel penetration sleeves, dissimilar metalwelds, bellows, and steel components that are subject to cyclic loading, but have nocurrent licensing basis fatigue analysis. The enhancement will be implemented prior tothe period of extended operation. LRA Appendix A, "Updated Safety Analysis ReportSupplement," and Appendix B, "Aging Management Programs," Sections A.1.22 and
 
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2L-11-238Page 2 of 15B.2.22, both titled "Inservice Inspection (ISI) Program -IWE," and Table A-I,"Davis-Besse License Renewal Commitments," are revised to include this enhancementand new license renewal future commitment.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI B.2.39-10Background:By letter dated May 24, 2011, the applicant responded to a staff RAIs B.2.25-4 andB.2.39-2 regarding operating experience with spent fuel pool (SFP) leakage. In theresponse the applicant stated that based on visual inspections of walls or floorsadjacent to the SFP the current leakage appears to be contained within the leakchase channels. The applicant also committed (Commitment 37) to take corebores prior to the period of extended operation of the two locations whereleakage had been identified.Issue:1. The applicant stated that based on visual inspections, the current leakageappears to be contained within the leak chase channels; however, theapplicant did not discuss an increased inspection frequency to continueto verify this, nor did the applicant commit to keeping the leak chasechannels and unlined leak trenches clear. The applicant also stated thatfor small leaks visual observations are insufficient to monitor the leakagestatus.2. The applicant reported that more comprehensive chemical analyses wereperformed in the past (e.g., as that carried out in 1996); however, theapplicant did not discuss plans to perform comprehensive analysesduring the period of extended operation. Chemical analyses performed ina timely fashion for effluent acidity and iron content (e.g., for pH, monthly;iron, semiannually) will assure that the effluent is not contributing toconcrete degradation and corrosion of steel leak chase members andrebars. Thus the activity assures the integrity of the spent fuel poolremains during the period of extended operation.3. Although the applicant committed to taking core bores, the response didnot provide details about when the cores would be taken, or why theexisting condition was acceptable until cores are taken. The responsealso did not address what would be included in the evaluation of thecores, or the acceptance criteria that would be used in the evaluation.
 
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2L-1 1-238Page 3 of 154. Minimal information was provided on the evaluation that will be done onthe underside of the SFP (Commitment 38).Request:1. Explain how leakage will be kept from migrating through the concretewalls. If keeping the leak chase channels clear will be part of the solution,explain how the channels will be kept clear. If this involves a reoccurringactivity, justify the frequency between occurrences. Also discuss andprovide technical justification whether or not more frequent visualinspections that could include the use of boroscopes will be conductedon the SFP leak chase system assuring its functionality and on thesurrounding concrete to verify that overflow leakage from cloggedchannels is not migrating through the surrounding concrete.2. Provide more details as to why there is no frequent chemical analysis ofthe collected leakage to assure its aciditylpH remains comparable to thatof the pool and its iron content is minimal.3. Provide more details about when the concrete cores will be taken,including the frequency and timing to establish a trend. Explain whatevaluations will be done on the cores and what criteria will be used todetermine the adequacy of the effected concrete.4. Provide more information on the evaluation discussed in Commitment 38,including possible actions and a preliminary schedule. Provide thecriteria for determining the need to repair the cracking located on theunderside of the SFPRESPONSE RAI B.2.39-101. Leakage will be kept from migrating through the concrete walls of the spent fuelpool, the transfer pit and the cask pit by allowing leakage to continuously drainthrough the leak chase piping, and by keeping the leak chase piping and valvesclear. Those activities also include the prevention of leakage through the floorsof the spent fuel pool, the transfer pit and the cask pit. FENOC will performmore frequent inspections of the inside of the leak chase piping and of theoutside of the walls and floors (from the ceiling side) where those areas areaccessible. Also, FENOC currently monitors the inaccessible areas south of theAuxiliary Building wall as part of the FENOC implementation of the NuclearEnergy Institute Groundwater Protection Initiative for Davis-Besse, asdescribed in the response to RAI B.2.25-7.With respect to the vertical and horizontal leak chase channels, the leak chasetrenches, the horizontal leakage collector tubes and the leak chase piping, theinstalled configurations of the spent fuel pool, the transfer pit and the cask pit
 
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2L-1 1-238Page 4 of 15are the same except for the lower 6.5 feet of the cask pit. The liner walls of thelower 6.5 feet of the cask pit were not used as formwork for pouring theconcrete walls of the cask pit. All of the other liner walls were used as formworkfor pouring the associated concrete walls. Due to the configuration of the leakchase channels, the leak chase trenches, the leakage collector tubes and theleak chase piping, it is not practicable to verify by 100% visual inspection thatthe leak chase channels and trenches are clear. The leak chase channels andcollector tubes (with the exception of the lower 6.5 feet of the cask pit) areembedded in concrete because they were welded to the spent fuel pool linerwalls before the liner walls were used as the formwork for pouring the concretespent fuel pool walls. The concrete trenches were formed in the concrete floorsof the spent fuel pool, transfer pit and cask pit structures and then the linerfloors were placed directly on top of the concrete floors. The 21 leak chasezone drain pipes connect to the bottoms of the trenches and to the horizontalleakage collector tubes that are located at the bottoms of the pool and pits. Theonly possible access to the leak chase trenches is from below through the leakchase piping. The drain pipes connect at the bottoms of the leak chasetrenches at a ninety-degree angle. All drain pipes but two have at least twoninety-degree elbows. Even by cutting the pipes at the ceiling (floor of the poolsand pits) access for a boroscope would be limited by the required ninety-degreeturns. The only possible access to the leakage collector tubes is from belowthrough the leak chase piping or from above through the vertical leak chasechannels. Attempts to view the leakage collector tubes from below would beimpeded by the bends in the leak chase piping and the ninety-degree turnsrequired for the horizontal entry to the leakage collector tubes. Attempting toview the leakage collector tubes from above is not practicable due to thesignificant foreign material exclusion concerns that would be introduced by anattempt to gain access the tops of the leak chase channels. Access from abovewould also be impeded by the need to make ninety-degree turns with aboroscope or other video equipment at the bottoms of the vertical leakchase channels.Therefore, more frequent inspection of the leak chase zone drain piping will beconducted. A new reoccurring preventive maintenance (PM) activity has beendeveloped to inspect and clean the zone drain piping and valves. FENOC willperform the PM every 18 months beginning prior to entering the period ofextended operation. The PM is modeled on a maintenance order that has beenperformed infrequently in the past to inspect and clean the zone drain pipingand valves. The frequency of inspecting and cleaning the leak chase pipingpathways is based on plant operating experience. For example, in 2007 whenthe drain piping and valves were cleaned, only one line was found to beclogged. No migration of water through walls has been identified since 2001.However, in 2007, leakage was identified coming from the ceiling of Room 109,below the spent fuel pool, about 20 months after the last inspection andcleaning of the drain piping. After the 2007 inspection and cleaning of the drainlines, no migration of water through the walls or ceiling was indentified until
 
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2L-1 1-238Page 5 of 152011, about 42 months after the inspection and cleaning of the drain lines.Therefore, the 18-month frequency of the cleaning and inspection isappropriate. In order to further assure the adequacy of the 18-month frequency,FENOC will inspect the accessible outside walls and floor (from the ceiling side)of the pool and pits once a year. This inspection will be a documentedinspection performed with the specific intent of identifying indications of leakagemigrating through the walls.In addition to the inspections and cleaning, current practice is to maintain theleakage pathway valves open for those pathways that have the most leakage.This practice allows for enough flow to keep the pathways clear. The totalleakage is very small, typically about 2 milliliters/minute and is divided amongthree to seven of the twenty-one available leak chase pathways. Typically,keeping one or two pathways open would drain most of the leakage.2. In lieu of providing more details as to why there is no frequent chemical analysisof the collected leakage, FENOC commits to conduct more frequent chemicalanalysis of spent fuel pool leakage. In the past, for the drainage from the leakchase piping, only boron concentration was measured. FENOC will analyzecollected leak chase drainage for pH monthly, and for iron every six months.Because there is no history of chemical analyses for parameters other thanboron, the acceptance criteria may have to be adjusted in the future to reflectactual plant-specific operating experience with the analyses. The initialacceptance criteria will be 7.0 to 8.0 for pH. The results for iron will be monitoredand trended to insure that there is no indication of corrosion of the reinforcingbars in the walls or floor of the pool and pits. An acceptance criterion for the ironanalyses will be developed after three years of measurements.3. As stated in the May 24, 2011 FENOC response (MLI 11151A090) to RAIB.2.39-2, FENOC will perform core bores of the ECCS Pump Room No. 1 walland the Room 109 ceiling. The core bores will be deep enough to exposereinforcing bar in the wall and ceiling. The core samples from the core boreswill be examined for signs of corrosion or chemical effects of boric acid on theconcrete or reinforcing bars. The examination will include a petrographicexamination. The reinforcing steel that will be exposed for a visual inspectionwill have corrosion products collected for testing. Degradation identified fromthe samples will be entered into the FENOC Corrective Action Program. Thecore bores will be performed in areas where leakage has been observed in thepast. The first set of core bores will be performed by the end of 2014. Thesecond set of core bores will be performed by the end of 2020. Further corebores will be conducted, if warranted, based on the evaluation of the results ofthe inspection and testing of the core bores or if SFP leakage through the wallor ceiling recurs after the second set of core bores is performed. If SFP leakagethrough another wall or ceiling is identified, then core bores will be performed ina manner similar to that stated for the ECCS Pump Room No. 1 wall and theRoom 109 ceiling.
 
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2L-1 1-238Page 6 of 154. As noted in the response to part 3 of this RAI, a core bore of the Room 109ceiling will be performed by the end of 2014. Degradation identified from thesamples will be entered into the FENOC Corrective Action Program. Thecondition of the concrete and the reinforcing steel will be evaluated at that timeto assist in determining what repairs, if any, need to be made to the undersideof the spent fuel pool concrete. The criterion for determining the need to repairthe cracking will be the continued capability of the structures to perform theirintended functions during the period of extended operation.Plant-specific operating experience has shown that the effect of borated wateron concrete and reinforcing steel has been relatively minor. A report preparedby Sargent & Lundy, SL-008105, evaluated the condition of concrete insidecontainment that had been exposed to borated water from refueling canalleakage. That report determined that even though boric acid had begun to reactwith the concrete at the cracks through which it had flowed, there was noindication that the amount of scale formed was sufficient to have significantlydegraded the concrete through which the boric acid had flowed. The reportdescribes the investigation, including the destructive testing (i.e., core bores),used to evaluate the condition of the refueling canal concrete and reinforcingsteel. The report identifies areas of the refueling canal concrete that displayedrust-stained cracks. Those cracks were investigated and it was determined thatthe reinforcing steel had not been adversely impacted.The Davis-Besse plant-specific operating experience has been confirmed bysignificant industry information that was discussed in the Salem LicenseRenewal Safety Evaluation Report (SER) (ML110900295), Section 3.0.3.2.15,and the Prairie Island License Renewal SER (ML092890209), Section3.0.3.2.17. The evaluation of the core bores described in the response to part 3of this RAI will determine whether or not the Davis-Besse SFP leakage hasaffected the concrete and reinforcing steel in a manner that is not bounded bythe industry and Davis-Besse current operating experience.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI B.2.40-2BackgroundBy letter dated May 24, 2011, the applicant responded to a staff RAI B.2.40-1regarding operating experience with degradation of the north embankment of thesafety-related portion of the intake canal. In the response the applicant stated thata preventive maintenance has been initiated to monitor the embankment for any
 
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2L-1 1-238Page 7 of 15changes, both above and below the water. The applicant also stated that long-term plans have been developed for further evaluation of the embankment.Issue:Although the applicant stated long-term evaluation plans had been developed,they did not commit to completing the investigation and possible repairs prior tothe period of extended operation.Request:Commit to completing the investigation, and possible repairs, of the safety-related intake canal embankment prior to the period of extended operation, orexplain why it is unnecessary.RESPONSE RAI B.2.40-2The investigation of the existing degradation of the north embankment of thesafety-related portion of the intake canal is currently in progress. FENOC provides anew license renewal future commitment in LRA Table A-1, "Davis-Besse LicenseRenewal Commitments," to ensure that the investigation is completed prior to the periodof extended operation. Upon completion of the investigation, FENOC will evaluate theresults and complete needed repairs or modification prior to the period of extendedoperation.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI 3.5.2.3.12-3Background:By letter dated June 3, 2011, the applicant responded to a staff RAI 3.5.2.3.12-1regarding steel restraints in a backfill environment. The applicant stated that lossof material was not an applicable aging effect and referenced a study related tosteel piles in undisturbed soil. The applicant also stated that opportunisticinspections would be conducted of the steel restraints if excavation workuncovers the components.Issue:The staff does not agree that the referenced study regarding piles in undisturbedsoil applies to steel in structural backfill. Undisturbed soil has low oxygen levelswhich may limit corrosion. These conditions may not be present in structural
 
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2L-1 1-238Page 8 of 15backfill. In addition, the portion of Commitment 20 discussing opportunisticinspections only mentions concrete components.Request:1. Explain why the referenced study is applicable to the steel restraints inbackfill, or propose an appropriate aging management program tomanage loss of material for the steel restraints. If the proposal involvesfocused inspections, justify the adequacy of the inspection technique andfrequency.2. Explain whether or not the opportunistic inspections apply tocomponents other than concrete, and update the commitment asnecessary.RESPONSE RAI 3.5.2.3.12-31. The steel restraints are steel cables that were used to restrain theEmergency Diesel Generator (EDG) fuel oil storage tanks (week tanks)during plant construction before backfill was placed over the tanks. Onceconstruction was completed, the steel restraints for the EDG week tanks wereno longer needed. Therefore the steel restraints have no current licensing basisfunction and should not have been in scope for license renewal. The LRA isrevised to remove the steel EDG Fuel Oil Storage Tank Hold Down Restraintsfrom scope; no aging management program is needed to manage loss ofmaterial for the steel restraints.2. Because the steel restraints in backfill are not in scope for license renewal,there is no need for opportunistic inspections of the steel restraints.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI B.2.25-7Background:By letter dated May 24, 2011, the applicant responded to a staff RAI B.2.25-4regarding the selection of the monthly leakage rate inspection and pointed outthat leakages vary from site to site and that there is not industry standard on themonitoring frequency. The applicant further determined that the Davis-Bessefrequency of monthly leakage collection is sufficient for monitoring long termchanges to the liner and leak chase system.
 
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2L-1 1-238Page 9 of 15Issue:The staff notes that even though operating experience may vary withconfiguration and status, nevertheless, there are examples of plants with similarleakage rates as DBNPS that perform the task on a daily basis. The staff alsonotes that the applicant collects 30 gallons of effluent per month from the leakchase system (a gallon a day) the units of the acceptance criteria, however, are inmilliliters per minute. The staff further notes that according to IN 2004-05leakages, if not identified in a timely fashion, could potentially have detrimentaleffects to SSCs and the environment.Request:1. Explain the discrepancy of units of the collected effluent gallons perday/month vs. the acceptance criteria units of milliliters per minute. Arethe milliliters per minute used during the monthly collections of theleakage or these are just averaging units reducing the gallons perday/month to milliliters per minute?2. Identify any actions taken subsequent to the issue of IN 2004-05 anddiscuss if the leakage rates in excess of 15 milliliters per minute, stated inCommitment 30, would be considered critical enough to consider morefrequent monitoring.RESPONSE RAI B.2.25-71. There is no discrepancy of units in the collected effluent gallons per day/monthversus the acceptance criteria units of milliliters per minute. In the response toRAI B.2.25-4, FENOC stated that, "The leak rate from the spent fuel poolmonitoring channels is very small, typically ranging from zero to a few millilitersper minute (on the order of one gallon per day)." FENOC included the roughconversion of milliliters per minute to "on the order of one gallon per day" toprovide a comparison of the relative sizes of the measured leak rate and thevolume of the spent fuel pool. In the response to RAI B.2.25-4, FENOC did notinclude an estimate of the volume of effluent collected per day or per month.The monthly monitoring of leakage is accomplished by collecting andmeasuring milliliters of leakage in a beaker and then computing a leak rate inmilliliters per minute, based on the time recorded for collection of the leakage.The collection time for each zone is at least 40 minutes. There is no directmeasurement of gallons per day or per month.2. Subsequent to the issue of NRC Information Notice (IN) 2004-05, datedMarch 3, 2004, "Spent Fuel Pool Leakage to Onsite Groundwater"; FENOCinitiated Condition Report (CR) 04-01719, dated March 5, 2004, "IN 2004-05Spent Fuel Pool Leakage to Onsite Groundwater." The purpose of
 
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2L-1 1-238Page 10 of 15CR 04-01719 was to evaluate IN 2004-05 for similar concerns at Davis-Besse.The evaluation determined that IN 2004-05 describes the same incident at theSalem Nuclear Generating Station that had been described in an Institute ofNuclear Power Operations (INPO) Operating Experience (OE) Report.IN 2004-05 noted that tritium had been detected in groundwater in two testlocations near the Salem Unit 1 Fuel Handling Building. FENOC had initiatedCR 03-02360 on March 25, 2003, for the evaluation of the INPO OE Report.The Investigation Summary for CR 04-01719 determined that the correctiveactions specified in CR 03-02360 had adequately addressed the issuedescribed in IN 2004-05. Two corrective actions had been specified forCR 03-02360. The first corrective action was to verify that the leak collectionisolation valves were not clogged with boric acid. That corrective actionincluded necessary cleaning or replacement of the isolation valves. The workorder for that corrective action was completed in 2005. FENOC initiated CR 07-13318 on January 24, 2007, due to boric acid observed on the ceiling of Room109, which is below the spent fuel pool. Corrective Action #1 for CR 07-13318to unclog the tell-tail drains to ensure that leakage from the Spent Fuel Pool willnot leak onto the ceiling on the 545 foot elevation was completed on August 31,2007. Valve SF99-T was found to be clogged and was cleaned with a steamcleaner and stiff wire. The other valves and the associated tell-tale drains werenot clogged in the area that the work order addressed.FENOC initiated CR 11-90368 due to boric acid observed on the ceiling ofRoom 109. In 2011, as corrective action for CR 11-90368, the leak monitoringlines were checked for clogging. Two lines were found to be clogged and theywere cleared. Also, as corrective action for CR 11-90368, a preventivemaintenance (PM) activity was created to verify that the leak monitoring linesare not clogged and to clean the lines as needed. The frequency of the PM isonce every three years.The second corrective action for CR 03-02360 was to obtain the appropriatenumber of soil samples to confirm whether there was evidence of contaminationin the soil due to leakage of the Spent Fuel Pool or Cask Pit. The south walls ofthe Spent Fuel Pool and the Cask Pit form the outside wall of the AuxiliaryBuilding. An existing monitoring well about 75 feet from the south wall of theAuxiliary Building was chosen for testing. That well was considered to be an ideallocation as there would be a minimal travel distance for any possible tritiatedwater. Based on the results of the testing, and in comparison with other routinesample results, there was no evidence to indicate that there was or had beenleakage from either the Spent Fuel Pool or the Cask Pit to the environment. TheCorrective Actions for CR 03-02360 were summarized in the InvestigationSummary for CR 04-01719. Based on FENOC management review of CR 04-01719, an activity was created to require periodic sampling of the groundwaterfrom the monitoring well (MW-1 8) nearest the Spent Fuel Pool for tritiumconcentration. To date, Monitoring Well MW-18 has been sampled four more
 
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2L-1 1-238Page 11 of 15times since the July 28, 2004 sampling in response to CR 03-02360. All of thesamples had tritium concentrations lower than observed from the July 28, 2004,sampling. Background tritium levels have been statistically determined by up-gradient groundwater sampling and sampling of Lake Erie waters to be between178 and 348 picoCuries per liter (pCi/L). The highest of the recent Well MW-1 8sample results was 436 pCi/L. The other recent samples have been below the348 pCi/L background level. Since 2009, the sample results have been reportedin the "Onsite Groundwater Monitoring Section" of the Combined AnnualRadiological Environmental Operating Report and Radiological Effluent ReleaseReport for the Davis-Besse Nuclear Power Station. The minimum samplingfrequency for Well MW-1 8 is once every five years.Many other onsite wells are sampled for radionuclides, including tritium, as partof the FENOC implementation of the Nuclear Energy Institute GroundwaterProtection Initiative for Davis-Besse. For example, 16 new monitoring wells wereinstalled in August 2007 in six distinct locations. The installation of those wells isdescribed in the Groundwater Monitoring Well Installation and Monitoring Reportdeveloped by Engineering Resources Management (ERM) on behalf of FENOC.The ERM report, completed on March 18, 2008, also notes that the Spent FuelPool, Fuel Transfer Canal and Cask Pit are potential sources of elevated tritiumdetected in groundwater due to past instances of leakage.If the leakage rates in excess of 15 milliliters per minute, stated in licenserenewal future Commitment 30, were to be exceeded, such a change would beconsidered critical enough to consider more frequent monitoring. Commitment 30is revised to include consideration of more frequent monitoring.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI B.2.25-8Background:By letter dated May 24, 2011, the applicant responded to a staff RAI B.2.25-6regarding the Plant-Specific Leak Chase Program USAR supplement. Accordingto the applicant the description provided in A 1.25, is consistent with the SRP-LRrecommendations to provide the basis for determining that aging of the liners willbe managed, and briefly describe the program activities (i.e., leakage monitoring).Issue:The staff notes that the USAR supplement needs to be more descriptiveincorporating in its description the applicant's acknowledgement in its response
 
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2L-1 1-238Page 12 of 15to RAI B.2.25-6 that the program manages loss of material for the spent fuel pool,the fuel transfer pit, and the cask pit stainless steel liners.Request:Please update the USAR supplement to appropriately reflect the material,environment, and aging effect the program manages for the spent fuel pool, thefuel transfer pit, and the cask pit liners.RESPONSE RAI B.2.25-8LRA Sections A.1.25 and B.2.25, both titled, "Leak Chase Monitoring Program," arerevised to reflect the material, environment, and aging effect the program manages forthe spent fuel pool, the fuel transfer pit, and the cask pit liners.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI 2.3.3.18-3Background:LRA Section 2.3.3.18, "Makeup and Purification System," states that the letdowncoolers, designated as DB-E25-1 and -2, are not subject to aging managementreview because these components are periodically replaced and evaluated asshort-lived components. Since these are normally long-lived passive componentssubject to aging management review, the staff issued RAI 2.3.3.18-2 requestingthe basis for the replacement frequency and the circumstances surrounding theneed to replace these heat exchangers.In its response dated June 3, 2011, Davis-Besse stated that the coolerreplacement frequency is based on a qualified life from plant-specific operatingexperience, and is scheduled approximately every 14 years. The response alsostated that the cooler design "has a tendency to develop leaks" after 14 to 16years. The response further stated that the need to replace the coolers wasattributed to fatigue cracking due to flow-induced vibration, and that an extent ofcondition review determined that the design of these coolers is unique and noother similar heat exchangers are installed at Davis-Besse.Issue:As previously noted in RAI 2.3.3.18-2, if the frequency is based on qualified life,then information should be provided to demonstrate that the cooler's intended
 
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2L-1 1-238Page 13 of 15function is being maintained consistent with the current licensing basis, at thepoint in time immediately prior to replacement. The staff notes that in accordancewith SRP LR Section A.1.2.3.4, an aging management approach based solely ondetecting component failures is not considered an effective program. The staffalso notes that in accordance with USAR Section 3.9.2, and Table 3.9-2, theletdown coolers are safety related components constructed to the ASME Code,Section III, Class 3.In addition, the staff notes that, if the design of the cooler results in "a tendencyto develop leaks after.. .14 to 16 years," then each heat exchanger would haveonly been replaced twice, so far, at Davis-Besse. With the relatively limitedoperating experience and the limited number of data points, the ability toreasonably predict the life of the coolers appears to have a large degree ofuncertainty. In addition, as noted in RAI 2.3.3.18-2, previous LRAs for other siteshave attributed to the fatigue cracking problem in these letdown coolers to beassociated with specific operational transients, and, if a similar phenomenon isoccurring at Davis-Besse, then a predicted life may need to consider transients inaddition to operational time.Request:1. Provide a summary of Davis-Besse's operating experience associatedwith the letdown coolers, including occurrences of tube leakage and pastreplacements for each cooler. Consider including the circumstances howthe associated leakage from the reactor coolant system into thecomponent cooling water system was detected, and the approximatemagnitude(s) of the leakage.2. Provide a summary of any past evaluations of the cause(s) for previoustube leakage, including how leakage was determined to be from fatiguecracks due to flow-induced vibration, and the degree and extent of thecracking identified. Include information regarding the role any operationaltransients may have played in causing previous tube leakage or how itwas concluded that operational transients need not be considered.3. Provide the information that determined the cooler's intended function isbeing maintained consistent with current licensing basis, at the point intime immediately prior to replacement.RESPONSE RAI 2.3.3.18-31. In the 1980's, based upon experience at other Babcock & Wilcox (B&W)utilities, Davis-Besse participated with B&W and the other B&W utilities inefforts directed at the review and improvement of the reliability of the letdowncoolers. The evaluation reviewed the following areas of concern: stress
 
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2L-1 1-238Page 14 of 15considerations, flow induced vibration, tube bundle stress relief, secondary sidechemistry, system operational considerations and tube bundle stitch welds. Theevaluation reviewed letdown cooler operational considerations andrecommended parallel cooler operation. The Davis-Besse normal operatingconfiguration is parallel operation.The coolers performed satisfactorily from 1977 to 1991. In 1991 Davis-Besseexperienced high contamination levels in the Component Cooling Water (CCW)System Pump Room, and indications of reactor coolant leakage of 13 to 30gallons per day into the CCW System. Troubleshooting was performed whichidentified a tube leak in the #1 letdown cooler (E25-1). Both coolers werereplaced in 1993.On December 30, 2009 chemistry gamma analysis detected short half-lifeisotopes in the CCW sample. Indications showed this was a small activeReactor Coolant System (RCS) leak (less than one gallon per day). Chemistryperforms CCW samples on a weekly basis. The chemistry sample point is fromthe in-service CCW heat exchanger. Both letdown coolers were replaced in2010. A fixed interval replacement preventive maintenance task was created.2. Industry and site operating experience indicate that, based on the design of thecoolers, the letdown coolers have a tendency to develop tube leaks. The B&Wreport regarding the reliability of letdown coolers identified the cause as fatiguecracking likely initiated by flow-induced vibration.The overall conclusion from an operating experience review regarding letdowncooler leakage identified similar issues across the industry and at Davis-Besse.Corrective Actions taken regarding leakage through the letdown coolersincluded plugging of affected tubes and/or replacement of the coolers. Theletdown cooler design at Davis-Besse contains 30 helical-type tubes which arenot designed to be plugged. The vendor does not support tube plugging in thistype of heat exchanger. Additionally there is no operating experience whichsupports successful plugging of this design in the industry. Therefore, theconclusion reached was to replace the letdown coolers.A failure analysis to determine the specific leak location/mechanism on theletdown cooler was not performed. Performing a failure analysis becomes aradiation dose concern due to the high dose rates associated with the coolers.No specific operational transients were identified that played a role in causingthe previous tube leakage. One notable transient (D1 and D2 Bus lockout)occurred on October 14, 1998, and is described in Davis-Besse Licensee EventReport (LER) 1998-011-00. The transient resulted in a water-hammer eventthat caused damage to a CCW letdown cooler overpressure-protection rupturedisk. Corrective actions implemented to reduce the consequences of future'loss of CCW pump' transients included replacement of the rupture disks with a
 
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2L-1 1-238Page 15 of 15relief valve. However, it can not be concluded that the water-hammer eventcontributed to the leaks in the letdown coolers since the leaks began more than10 years after the event occurred.3. A Problem Solving Team was formed in January 2010, when it was identifiedthat RCS activity was present in the CCW System based on Chemistrysamples. An Operational Decision Making Issue (ODMI) was initiated formonitoring and trending various parameters with specific trigger points. Lowlevel conservative trigger points were established in the ODMI for the monitoredparameters. No trigger points were reached during the monitoring period.Based upon CCW activity levels remaining very low, RCS unidentified leakageessentially unchanged, less than the 0.1 gallons per minute trigger established,absence of radiation monitoring alarms, and no unexplained increases in CCWSurge Tank level, the cooler's intended function was maintained consistent withcurrent licensing basis prior to replacement of the coolers.
 
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3L-1 1-238Reply to Request for Additional Information for the Review of theDavis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC letter dated July 21, 2011 (MLI 11196A1 27)Page 1 of 7Question RAI B.2.41-1Background:The SRP-LR, Revision 2, Section A.1.2.3.4, describes the recommendations for anacceptable detection of aging effects program element for a plant-specificprogram and states that "the discussion [of the inspection method or technique]should provide justification, including codes and standards referenced, that thetechnique and frequency are adequate to detect the aging effects before a loss ofSC-intended function."FirstEnergy Operating Company's (the applicant) Inspection of Internal Surfacesin Miscellaneous Piping and Ducting Program, as provided in License RenewalApplication (LRA) Section B.2.41 by letters dated May 24, 2011, and June 3, 2011,states that "enhanced visual exams" will be conducted to manage cracking forsusceptible stainless steel components. However, the LRA does not state to whatpredetermined criteria (standard) enhanced visual examinations will beconducted.Issue:The LRA does not state the standard that enhanced visual examinations will beconducted against.Request:Revise LRA Section B.2.41, Inspection of Internal Surfaces in MiscellaneousPiping and Ducting Program, to indicate the standard to which enhanced visualexaminations will be conducted in order to manage cracking.RESPONSE RAI B.2.41-1When required by the ASME Code, inspections are conducted in accordance with theapplicable code requirements. In the absence of applicable code requirements, visualinspections are performed of metallic and polymeric component surfaces usingplant-specific procedures implemented by inspectors qualified through plant-specificprograms. LRA Sections A.1.41 and B.2.41, both titled "Inspection of Internal Surfacesin Miscellaneous Piping and Ducting Program," are revised to include this information.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
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3L-11-238Page 2 of 7Question RAI B.2.41-2Background:The SRP-LR, Revision 2, Section A.1.2.3.10, describes the recommendations foran acceptable operating experience program element for new aging managementprograms (AMPs) and states that "an applicant should commit to a review offuture plant-specific and industry operating experience for new programs toconfirm their effectiveness."LRA Table A-I, Davis-Besse License Renewal Commitments, does not include acommitment to perform a review of future operating experience to confirm theeffectiveness of the new Inspection of Internal Surfaces in Miscellaneous Pipingand Ducting Components Program.Issue:This program's LRA commitment list is not consistent with the current staffposition as stated within the SRP-LR, Revision 2, concerning reviews of futureoperating experience for new aging management programs.Request:Revise LRA Table A-I, Davis-Besse License Renewal Commitments, for theInspection of Internal Surfaces in Miscellaneous Piping and Ducting ComponentsProgram to include a commitment to perform a future review of operatingexperience to confirm the effectiveness of this program or justify why such areview is not necessary.RESPONSE RAI B.2.41-2By letter (ML11180A060) dated June 24, 2011, in response to RAI B.1.4-1, FENOCprovided license renewal future commitment number 43 to "[e]nsure that the currentstation operating experience review process includes future reviews of plant-specificand industry operating experience to confirm the effectiveness of the license renewalaging management programs, to determine the need for programs to be enhanced, orindicate a need to develop new aging management programs." Therefore, a separateoperating experience commitment for the Inspection of Internal Surfaces inMiscellaneous Piping and Ducting Components Program is not necessary.
 
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3L-11-238Page 3 of 7Question RAI B.2.41-3Background:SRP-LR, Revision 2, Table 3.0-1 states that the recommended description for thefinal safety analysis report (FSAR) supplement for a plant-specific AMP shouldinclude the bases for determining that aging effects will be managed during theperiod of extended operation.The applicant's Inspection of Internal Surfaces in Miscellaneous Piping andDucting Program's updated safety analysis report (USAR) supplement provided inLRA Section A.1.41 by letters dated May 24, 2011, and June 3, 2011, does not statewhat type of inspections will be used to manage the program's aging effects.Issue:The USAR supplement for the applicant's Inspection of Internal Surfaces inMiscellaneous Piping and Ducting Program does not state the type of inspectionsthat will be used to manage the program's aging effects, and therefore does notadequately describe the basis for how the program will manage aging effectsduring the period of extended operation.Request:Revise the USAR supplement associated with the Internal Surfaces inMiscellaneous Piping and Ducting Program to include the type of inspections thatwill be used to manage the program's aging effects, consistent with SRP-LR,Revision 2, or justify why the revision is not necessary.RESPONSE RAI B.2.41-3LRA Section A. 1.41, Inspection of Internal Surfaces in Miscellaneous Piping andDucting Program, is revised to include the following table showing the types ofinspections to be performed based on aging effect and aging mechanism:Parameters Monitored or InspectedAnd Aging Effect for Specific ComponentAging Aging Parameter InspectionEffect Mechanism Monitored Method "ILoss of Material Crevice Surface Condition, Visual (VT-i or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material Galvanic Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)
 
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3L-1 1-238Page 4 of 7Parameters Monitored or InspectedAnd Aging Effect for Specific Component, cont.Aging Aging Parameter InspectionEffect Mechanism Monitored Method (')Loss of Material General Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material MIC Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Pitting Corrosion Surface Condition, Visual (VT-1 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Erosion Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Reduction of Fouling Tube Fouling Visual (VT-3 or equivalent) orHeat Transfer Enhanced VT-1 for CASSCracking SCC or Cyclic Surface Condition, Enhanced Visual (EVT-1 or equivalent) orLoading Cracks Surface Examination (magnetic particle,I liquid penetrant, or Volumetric (RT or UT)(1) When required by the ASME Code, inspections are conducted in accordance with theapplicable code requirements. In the absence of applicable code requirements, visualinspections are performed of metallic and polymeric component surfaces using plant-specificprocedures implemented by inspectors qualified through plant-specific programs.See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI 4.6-1Background:In LRA Section 4.6.1, "Containment Vessel," the applicant states that:Analysis of 400 pressure cycles (from -25 to 120 psi) and 400 temperaturecycles (from 30°F to 1200F) were performed against the requirements of theAmerican Society of Mechanical Engineers Boiler and Pressure VesselCode, Section III, Paragraph N-415.1. The applicant also states that thevalues of 400 pressure and temperature cycles used to exclude fatigueanalyses will not be exceeded for 60 years of operation and thus the TLAAsassociated with the exclusion of the containment vessel from fatigueanalysis per ASME Section III, Paragraph N-415.1 will remain valid throughthe period of extended operation in accordance with 10 CFR 54.21(c)(1)(i)
 
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3L-1 1-238Page 5 of 7In LRA Section 4.6.2, "Permanent Canal Seal Plate", the applicant states that:The fatigue analysis of the permanent canal seal plate seal membraneshows that the maximum fatigue usage factor is based on 50 fullheatup/cooldown cycles.Issue:The staff reviewed LRA Section 4.6, 4.3.1, and the applicant's USAR Section 3.8and did not find the design basis information regarding:1. The total number of transients used to determine that requirements of afatigue waiver per Subparagraph N-415.1 were met for the containmentvessel2. The basis for the number of transients used in the original fatigueanalysis of the permanent canal seal plateThe staff needs more information to confirm that fatigue evaluations for thecontainment vessel will remain valid for the period of extended operation inaccordance with 10 CFR 54.21(c)(1). The staff also needs more information toverify the number of cycles used in the design of the permanent canal seal plate.Request:Provide the following information:1. A description of the original design basis used to determine thatrequirements of a fatigue waiver per Subparagraph N-415.1 were met forthe containment vessel.2. The basis for the LRA statement that the maximum fatigue usage factorfor the permanent canal seal plate is based on 50 cycles.RESPONSE RAI 4.6-1Response to request number 1:The fatigue waiver calculation for the containment vessel was performed in accordancewith N-415.1(a) through N-415.1(f) of the ASME Code, Section II1. This calculationverified the requirements of N-415.1 against 400 pressure cycles (from -25 to 20 psi)
 
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3L-1 1-238Page 6 of 7and 400 temperature cycles (from 30°F to 120°F) as follows:N-415.1(a)The number of times (including startup and shutdown) that the pressure will becycled from atmospheric pressure to operating pressure and back to atmosphericpressure must not exceed the number of cycles on Figure N-415(A) correspondingto an Sa value of 3 times Sm .3 Sr, is equal to 56,250 psi and from Figure N-415(A) the corresponding number ofcycles is equal to 1,800. The specified number of 400 pressure cycles is less thanthe 1,800 cycles from Figure N-415(A). Therefore, the condition in N-415.1(a) is met.N-415.1(b)Specified full range of pressure fluctuations may not exceed the quantity 1/3 xdesign pressure x Sa/Sm. Sa is the value from Figure N-415(A) for 400 cycles.1/3 x 36 x 125,000/18,750 = 80 psiSpecified full range of pressure fluctuations is 45 psi (-25 to 20 psi) and is less than80 psi. Therefore, the condition in N-415.1(b) is met.N-415.1(c)The temperature difference in degrees F between any two adjacent points duringnormal operation and during startup and shutdown must not exceed Sa/(2Ea).For a mean temperature of 700F, 120,000/2(27.9 x 106)(6.07 x 10-6) = 3580FTemperature cycle range of 90°F (from 30°F to 120°F) is less than 358°F. Therefore,the condition in N-415.1(c) is met.N-415.1(d)The temperature difference in degrees F between any two adjacent points does notchange during normal operation by more than Sa/(2Ea).For a mean temperature of 700F, 120,000/2(27.9 x 106)(6.07 x 10-6) = 358°FTemperature cycle range of 90°F (from 30°F to 1200F) is less than 3580F. Therefore,the condition in N-415.1(d) is met.N-415.1(e)During normal operation, components fabricated from materials of differing moduli ofelasticity and/or coefficients of thermal expansion may not fluctuate more thanSa/2[(E1a,) -(E2a2)].
 
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3L-1 1-238Page 7 of 7For 18-8 stainless steel jointed to low carbon steel the range is125,000/2[(9.11 )(27.4) -(6.07)(27.9)] = 780°FThe value of 780°F is greater than the 120°F specified. Therefore, the condition inN-415.1 (e) is met.N-415.1(f)The specified full range of mechanical loads must not result in stresses exceedingSa for 400 cycles.3 Sm is equal to 56,250 psi and is less than Sa of 125,000 psi for 400 cycles fromFigure N-415(A). Therefore, the condition in N-415.1(f) is met.The pressure cycle range used in the fatigue waiver evaluation is from -25 to 20 psi, fora full range pressure fluctuation of 45 psi. However, the possible full range pressurefluctuation is from -0.67 to 45 psig based on the containment vessel design allowablenegative pressure of -0.67 psig and the containment vessel pneumatic test pressure of45 psig (design pressure of 36 psig times 1.25). This adjusted full range pressurefluctuation of 45.67 psi is less than the 80 psi value determined in N-415.1(b), above.Therefore, the condition in N-415.1(b) is met. In addition, clarification is needed that the60-year projected cycles for plant heatup and cooldown are less than the 400 pressurecycles and 400 temperature cycles specified in the fatigue waiver analysis.See the Enclosure to this letter for the revision to the DBNPS LRA.Response to request number 2:The original reactor cavity seal plate required critical path time for installation andremoval during refueling outages. The original reactor cavity seal plate was replaced in2004 with a new design that allowed for permanent installation. The fatigue analysis ofthe permanent reactor cavity seal plate shows that the maximum fatigue usage factor isat the inner leg to the reactor vessel seal ledge weld. This maximum fatigue usagefactor is 1.2 and is based on 60 heatup/cooldown cycles and 50 operating basisearthquake (OBE) cycles. The ASME Code requires that the fatigue usage factor notexceed 1.0. To satisfy this requirement, the allowable for the heatup/cooldown cycles isestablished at 50 cycles. The OBE cycles have a negligible contribution to the fatigueusage factor.As shown in LRA Table 4.3-1 for Transient 31A, the permanent reactor cavity seal plateis projected to experience 51 heatup/cooldown cycles by the end of the period ofextended operation. In addition, as shown in Table 4.3-1 for Transient 31 B, theassumption of 50 OBE cycles remains valid for the period of extended operation. Sincethe heatup/cooldown cycles are projected to exceed the allowable of 50 cycles by theend of the period of extended operation, the effects of fatigue will be managed by theFatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).
 
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4L-1 1-238Reply to Request for Additional Information for the Review of theDavis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Letter dated July 27, 2011 (ML 1203A080)Page 1 of 5Question RAI B.2.3-5Backaround:In response to a U.S. Nuclear Regulatory Commission (NRC or the staff) Requestfor Additional Information (RAI) B.2.3-2, in its letter dated June 3, 2011,FirstEnergy Nuclear Operating Company (the applicant) stated that periodictesting of contaminants by the Air Quality Monitoring Program is performed eachyear, and that the frequency of testing is based on the recommendations of theInstitute of Nuclear Power Operations (INPO) Supplemental Operating ExperienceReport (SOER) 88-1, "Instrument Air System Failures," which recommendsperiodic monitoring of air quality at several points throughout the instrument airsystem. Additionally, the SOER recommends that system air quality, as measuredat the discharge of the air dryers and after filters, should be maintained within therequirements of ANSI Standard ISA-S7.3. ANSI Standard ISA-S7.3 has beenwithdrawn and replaced with ANSI/ISA-7.0.01-1996. While ISA-$7.3 did not specifya frequency for checking the dew point, ANSI/ISA-7.0.01-1996 recommends pershift monitoring of the dew point if a monitored alarm is not available.Generic Aging Lessons Learned (GALL) Aging Management Program (AMP),XI.M24, "Compressed Air Monitoring," under the "Detection of Aging Effects,"program element states:[t]he program periodically samples and tests the air quality in thecompressed system for moisture in accordance with industry standards,such as ANSI/ISAS7.0.01-1996. Typically, compressed systems have in-linedew point instrumentation that either checks continuously using anautomatic alarm system or is checked at least daily to ensure that moisturecontent is within specifications.Issue:If the dew point is not maintained well below the system operating temperature (atleast 18 degrees F below the minimum local ambient temperature), condensationcould occur. As recommended in the GALL Report, steel and stainless steelpiping, piping components, and piping elements exposed to condensation shouldbe managed for loss of material, and the Compressed Air Monitoring Program isan acceptable program to manage aging.Based on the current industry standard and recommendations in the GALLReport, it is not clear to the staff how periodic testing once a year ensures that the
 
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4L-1 1-238Page 2 of 5dew point is maintained well below the system operating temperature duringnormal operation, and that condensation is not occurring internally.Request:Justify how periodic testing once a year ensures that the dew point is maintainedwell below the system operating temperature during normal operation, as well asduring outages and maintenance, such that the environment remains "dry-air."RESPONSE RAI B.2.3-5Instrument Air is designed to have a dew point of 180F below the minimum local ambienttemperature at 100 psig. In response to Generic Letter 88-014, Davis-Besse committedto maintaining the Instrument Air System with a dew point of at least 350F below zero. AControl Room annunciator exists for Instrument Air Dryer Trouble, with one of theactuating devices being high moisture content in the desiccant.In addition to the periodic testing of contaminants performed each year, monthly dewpoint readings are taken downstream of each of the air dryers.Question RAI B.2.28-1Background:License renewal application (LRA) Section B.2.28 describes the existingNickel-Alloy Management Program as plant-specific. The applicant states that theirprogram manages primary water stress corrosion cracking for nickel-alloypressure boundary components, other than reactor vessel closure head nozzlesand steam generator tubes, exposed to reactor coolant. The applicant notes thatNUREG-1801 Rev. 1, Section XI.M11, "Nickel-Alloy Nozzles and Penetrations,"does not contain program elements. The staff notes that in NUREG-1801 Rev. 2,Section XI.M1IB "Cracking of Nickel-Alloy Components and Loss of Material Dueto Boric Acid-Induced Corrosion in Reactor Coolant Pressure BoundaryComponents," a guideline for existing requirements has been defined for thisaging management program.Issue:The staff reviewed program elements one through six of the applicant's programagainst the acceptance criteria for the corresponding elements as stated inSRP-LR Section A.1.2.3, and has determined that the following information isneeded in order to complete its review of the applicants aging managementprogram.
 
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4L-1 1-238Page 3 of 5Request:On June 21, 2011, the NRC published a final rule, NRC-2008-0554, which, in part,established new requirements, under 10 CFR 50.55a(g)(6)(ii)(F), for the inspectionof American Society of Mechanical Engineer's Boiler and Pressure Vessel Code(ASME Code) Class 1 nickel-alloy butt welds in the reactor coolant pressureboundary.The staff requests that the applicant confirms incorporation of the requirements ofTitle 10 of the Code of Federal Regulations 50.55a(g)(6)(ii)(F), which implementsASME Code Case N-770-1, with certain conditions, into the applicant's Nickel-AlloyManagement Program.RESPONSE RAI B.2.28-1Davis-Besse has not implemented ASME Code Case N-770-1, "Alternative ExaminationRequirements and Acceptance Standards for Class 1 PWR Piping and Vessel NozzleButt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With orWithout Application of Listed Mitigation Activities, Section Xl, Division 1 ," at this time.As provided in 10 CFR 50.55a(g)(6)(ii)(F)(1), "Licensees of existing, operatingpressurized-water reactors as of July 21, 2011 shall implement the requirements ofASME Code Case N-770-1, subject to the conditions specified in paragraphs(g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(1 0) of this section, by the first refueling outage afterAugust 22, 2011." The first refueling outage after August 22, 2011 for Davis-Besse is theCycle 17 refueling outage that is presently scheduled for mid-year 2012.Therefore, the following enhancement will be implemented in the "Detection of AgingEffects" and "Monitoring and Trending" program elements for the Nickel-AlloyManagement Program prior to the period of extended operation:"Provide for inspection of dissimilar metal butt welds in accordance with therequirements of ASME Code Case N-770-1, "Alternative ExaminationRequirements and Acceptance Standards for Class 1 PWR Piping and VesselNozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld FillerMaterial With or Without Application of Listed Mitigation Activities, Section Xl,Division 1," as modified by the Code of Federal Regulations, 10 CFR 50.55aSection (g)(6)(ii)(F)."This enhancement is included as a new license renewal future commitment in LRAAppendix A, "Updated Safety Analysis Report Supplement," Table A-I, "Davis-BesseLicense Renewal Commitments."See the Enclosure to this letter for the revision to the DBNPS LRA.
 
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4L-1 1-238Page 4 of 5Question RAI B.2.36-5Background:The applicant's Selective Leaching Inspection is a new program that will beconsistent with GALL AMP XI.M33, "Selective Leaching of Materials." The SRP-LR,Revision 2, Section A.1.2.3.10 states that "for new programs, an applicant mayneed to consider the impact of relevant [operating experience] OE that resultsfrom the past implementation of its existing AMPs that are existing programs andthe impact of relevant generic [operating experience] OE on developing theprogram elements" and that "an applicant should commit to a review of futureplant-specific and industry operating experience for new programs to confirmtheir effectiveness."In its response to RAI B.2.36-3, dated May 24, 2011, the applicant stated that "Theprogram does not consist of ongoing activities, but will end upon completion ofone-time inspections of the sample set of components. Therefore, a futureconfirmation of program effectiveness is not applicable to the Selective LeachingInspection. If plant-specific operating experience indicates the potential forselective leaching after program completion, it will be addressed using theCorrective Action Program."Issue:Even though the Selective Leaching Inspection is a one-time inspection program,the results of the inspections and industry operating experience should bereviewed to assess the effectiveness of the program at identifying selectiveleaching prior to loss of component intended function. If any deficiencies areidentified, a review should be performed to determine whether the program shouldbe enhanced or a new program should be developed. It is unclear to the staff howoperating experience will be incorporated into the Selective Leaching Inspectionto confirm the effectiveness of the program.Request:State how future plant-specific and industry operating experience related to theSelective Leaching Inspection will be reviewed to confirm the effectiveness of theprogram, evaluate the need for the program to be enhanced, or indicate a need todevelop a new aging management program.RESPONSE RAI B.2.36-5By letter (ML11180A060) dated June 24, 2011, in response to RAI B.1.4-1, FENOCprovided license renewal future commitment number 43 to "[e]nsure that the currentstation operating experience review process includes future reviews of plant-specific andindustry operating experience to confirm the effectiveness of the license renewal aging
 
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4L-11-238Page 5 of 5management programs, to determine the need for programs to be enhanced, or indicatea need to develop new aging management programs." Therefore, a separate operatingexperience commitment for the Selective Leaching Inspection Program is not necessary.
 
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5L-1 1-238Reply to Request for Additional Information for the Review of theDavis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,from NRC Conference Calls and NRC Region III License Renewal InspectionPage 1 of 10Question RAI B.11.4-1 Supplemental Response -Operating Experience Statementin Appendix AThe NRC initiated a telephone conference call with FENOC on July 12, 2011, todiscuss the FENOC response to RAI B.1.4-1 (submitted in FENOC letter datedJune 24, 2011 (ML1118OA060)) on operating experience. The NRC staff stated thattheir preference is that a statement be added to License Renewal Application(LRA) Appendix A, "Updated Safety Analysis Report Supplement," regarding howFENOC plans to address the review and incorporation of operating experienceinto the license renewal aging management programs, not as a commitment, butrather as a separate discussion in the beginning paragraphs of LRA Appendix A.RESPONSE RAI B.1.4-1 Supplemental Response -Operating Experience Statement inAppendix ALRA Appendix A is revised to add the operating experience statement that waspreviously added to Appendix B in response to NRC RAI B.1.4-1 (ML11 180A060) to theentry paragraphs of LRA Appendix A.See the Enclosure to this letter for the revision to the DBNPS LRA.Question Supplemental Response -steam generator aging management reviewtube foulingThe NRC initiated a telephone conference call with FENOC on July 12, 2011, todiscuss primary side fouling of steam generator tubes. NRC stated that, during apublic meeting on February 18, 2011, industry Steam Generator Task Forcerepresentatives indicated that primary side fouling of steam generator tubes isnot an issue in the United States (refer to the meeting summary in ADAMS underaccession number ML110670317). In the applicant's license renewal application,reduction of heat transfer of the steam generator tubes in a borated reactorcoolant environment is addressed as an aging mechanism. Has there been anyinformation gained by the industry since the February 18, 2011, meeting thatwould suggest that primary side loss of heat transfer has become an issue? Ifthere is additional information, please provide it. If not, discuss your plans towithdraw the aging management review line item that deals with reduction in heat
 
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5L-1 1-238Page 2 of 10transfer of nickel alloy tubing and sleeves in a borated reactor coolantenvironment.RESPONSE Supplemental Response -steam generator aging management reviewtube foulingFENOC has not experienced reduction in heat transfer of the primary side of the nickelalloy tubing and sleeves for the Davis-Besse steam generators. In addition, FENOC isnot aware of any industry operating experience that would suggest that primary sideloss of heat transfer of the steam generator tubes in a borated reactor coolantenvironment has become an issue. Therefore, FENOC is withdrawing the agingmanagement review line items that are associated with reduction in heat transfer ofnickel alloy tubing and sleeves in a borated reactor coolant environment.Rows 30 and 31 of LRA Table 3.1.2-4, "Aging Management Review Results -SteamGenerators," are revised to show as "Not Used."See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI 4.3-17 Supplemental Response -Integrated Fen [footnote]The NRC initiated a telephone conference call with FENOC on July 12, 2011, todiscuss the FENOC response to RAI 4.3-17 (submitted in FENOC letter datedJune 17, 2011 (ML11172A389)). NRC indicated they had reviewed the FENOCresponse for RAI 4.3-17, and noted that the RAI response for the Surge LinePiping states, "Global Fen is calculated by dividing the Uen by the in-air CUF,"which is the same method as that stated in MRP-47. LRA page 4.3-28, andfootnote 2 in LRA Table 4.3-2, "Davis-Besse CUFs for NUREG/CR-6260Locations," state, "adjusted CUF is obtained by dividing the Uen by theglobal Fen."Original RAI 4.3-17 QuestionLRA Section 4.3.4.2 states that the surge line piping and high pressureinjection/makeup (HPI/MU) nozzle and safe end were evaluated using anintegrated Fen approach consistent with EPRI Technical Report MRP-47,"Guidelines for Addressing Fatigue Environmental Effects in a License RenewalApplication," Revision 1, Section 4.2.The staff noted that consistent with MRP-47, Section 4.2, the CUF and Uen arecomputed for each load pair and an effective Fen is calculated by dividing the Uenby the CUF. LRA Section 4.3.4 states that the maximum Uen is calculated with a
 
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5L-11-238Page 3 of 10global Fen and the adjusted CUF is then obtained by dividing the Uen by theglobal Fen.The staff noted that EPRI Technical Report MRP-47 has not been reviewed andapproved by the NRC. Furthermore, the applicant stated that in footnote 2 of LRATable 4.3-2 the global Fen is calculated using the method from Section 4.2 ofMRP-47. However, the term "global Fen" is not discussed in MRP-47. The stafffurther noted that the process of calculating global Fen is not discussed inthe LRA.Therefore, it is not clear to staff how the applicant determined theenvironmentally adjusted CUF for the surge line piping and HPI/MU nozzle andsafe end.The staff requests the following information:1. Justify that use of the integrated Fen approach in the EPRI MRP-47 isapplicable and adequately conservative to calculate Uen for the period ofextended operation.2. Clarify the term "global Fen" and how it is calculated for each component.Provide its relationship with Uen calculation methodology discussedin MRP-47.RESPONSE RAI 4.3-17 (Integrated Fen [footnote]) Supplemental ResponseBased on the information provided and the discussion of the topic with the NRC,FENOC agreed during the conference call that footnote 2 of LRA Table 4.3-2 shouldbe revised.Upon further review, FENOC determined that additional changes are required to LRASection 4.3.4.2, "Davis-Besse Evaluation," and Table 4.3-2, as related to theenvironmentally-assisted fatigue (EAF) evaluation results for the surge line piping.Footnote 2 of LRA Table 4.3-2 is revised to state that the adjusted CUF was calculatedusing 60-year projected cycles (except for best-estimate 60-year project cycles of 114used for heatup and cooldown events).Footnote 9 is added to LRA Table 4.3-2 to state that Fen was determined for eachtransient pair (integrated Fen approach), and the Uen for each pair is determined bymultiplying the in-air usage (adjusted CUF) for that transient pair by the Fen calculatedfor that pair. The Uen for each transient pair were summed to come up with cumulativeUen for that specific location. Fen presented in the table is a global Fen calculated bydividing the cumulative Uen by the adjusted CUF.
 
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5L-1 1-238Page 4 of 10LRA Section 4.3.4.2, subsection titled, "Surge Line Fatigue Results," is revised toindicate that the in-air CUF (adjusted CUF) values were based on 60-year projectedcycles (except for best-estimate 60-year project cycles of 114 used for heatup andcooldown events).See the Enclosure to this letter for the revision to the DBNPS LRA.Question Supplemental Response -steam generator aging management reviewtube-to-tubesheet weldThe NRC initiated a telephone conference call with FENOC on July 13, 2011, toobtain clarification on how FENOC manages cracking due to primary water stresscorrosion cracking (PWSCC) of the Davis-Besse steam generatortube-to-tubesheet welds in comparison with NUREG-1801 (Generic AgingLessons Learned (GALL) Report) and NUREG-1800 (Standard Review Plan forLicense Renewal).RESPONSE Supplemental Response -steam generator aging management reviewtube-to-tubesheet weldUpon further review after the conference call with the NRC, FENOC determined that thetube-to-tubesheet welds (Alloy 600 welds) for the Davis-Besse steam generators do nothave a license renewal intended function and therefore, are not subject to an agingmanagement review. The Davis-Besse steam generators are Babcock & WilcoxModel 177-FA, once through design. The tubes and the tubesheets of the steamgenerators form the pressure boundary between the fluid in the secondary system andthe reactor coolant system. As provided in USAR Section 5.5.2.3, the tubes areexpanded (to a partial depth) into the tubesheet and the tubes are seal welded to thetubesheet near the tube ends. The American Society of Mechanical Engineers (ASME)Boiler and Pressure Vessel (B&PV) Code, Section XI, Division 1, 1995 Edition with1996 Addenda, IWA-9000 defines a seal weld as a nonstructural weld intended toprevent leakage, where the strength is provided by a separate means. The "separatemeans" in this case being the tube-to-tubesheet expansion joint which forms thepressure boundary. The tube-to-tubesheet welds are seal welds and therefore, are notpart of the pressure boundary. No revision to the DBNPS LRA is required.
 
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5L-1 1-238Page 5 of 10Question RAI 4.1-2 (Code Case N-481) Revised ResponseThe NRC initiated a telephone conference call with FENOC on July 15, 2011, todiscuss the FENOC response to RAI 4.1-2 (submitted in FENOC letter datedJune 17, 2011 (ML11172A389)). NRC stated they did not agree with the entireresponse. In the response to RAI 4.1-2, FENOC stated that the flaw toleranceanalysis for the reactor coolant pump (RCP) casings that was used to support theASME Code Case N-481 alternate augmented visual inspection bases for the RCPcasings was not a time-limited aging analysis (TLAA). The NRC agrees that theJ-integral aspect of the analysis is not a TLAA. However, the NRC does not agreethat the cycle-dependent fatigue flaw growth analysis is not a TLAA. Although the2000 analyzed cycles excessively bound the 240 design cycles, the analysisindicates it is applicable to a 40-year design life.Original QuestionLRA Section 4.3.2.2.4 discusses the fatigue TLAA for the reactor coolant pump(RCP) casings and states that they were analyzed for fatigue by the OEM to meetthe requirements of the ASME Code Section III, 1968 Edition through Winter-1968Addenda. LRA Table 3.1.1 item 3.1.1-55 states that these pump casings will bemanaged by the applicant's Inservice Inspection Program.The applicant's licensing basis includes a flaw tolerance analysis for the RCPcasings that was used to support ASME Code Case N-481's alternate augmentedvisual inspection bases for the RCP casings. The staff noted that this flawtolerance analysis is documented in Structural Integrity Associates (SIA) TopicalReport No. SIR-99-040, Revision 1, "ASME Code Case N-481 of Davis-BesseReactor Coolant Pumps." (ADAMS Accession No. ML011200090, datedApril 23, 2001).The staff noted that the evaluation in Report No. SIR-99-040 includes acycle-dependent fatigue flaw growth analysis for the pump casings welds that isbased on a 40-year design life; however, the applicant did not identify thisanalysis as a TLAA.Justify why the fatigue flaw growth analysis for the RCP pump casing welds inSIA Topical Report No. SIR-99-040, Revision 1, does not need to be identified as aTLAA in accordance with 10 CFR 54.21(c)(1).RESPONSE RAI 4.1-2 (Code Case N-481) Revised ResponseDuring the telephone conference call, FENOC agreed with the NRC's position, andstated that the response to RAI 4.1-2 would be revised. The revised response isas follows:
 
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5L-1 1-238Page 6 of 10The reactor coolant pumps (RCPs) are the only ASME Code Class 1 pumps installed atDavis-Besse. The pump casings are constructed of cast austenitic stainless steel. Theapplicable ASME Code for the current Third Ten-Year Inspection Interval for Davis-Besse is ASME Section XI, 1995 Edition, through the 1996 Addenda, as modified by10 CFR 50.55a or relief granted in accordance with 10 CFR 50.55a. ExaminationCategory B-L-1 of this Code year requires volumetric examination on pump casingwelds. ASME Code Case N-481, "Alternative Examination Requirements for CastAustenitic Pump Casings," provides an alternative to the volumetric examinationrequirement. This code case allows the replacement of volumetric examinations ofprimary loop pump casings with fracture mechanics-based integrity evaluation (Item (d)of the code case) supplemented by specific visual examinations. Davis-Besse hasinvoked the use of Code Case N-481 in place of the volumetric examinationrequirements of Code Category B-L-1. The NRC has accepted Code Case N-481 foruse in inservice inspection programs.Code Case N-481 requires an evaluation to demonstrate the safety and serviceability ofthe pump casings. The evaluation for the Davis-Besse RCPs required by CodeCase N-481 is documented in Structural Integrity Associates (SIA) report SIR-99-040.This evaluation assumed a quarter thickness flaw, with length six times its depth, andshowed that the flaw will remain stable considering the stresses and material propertiesof the pump casing. To determine stability of the postulated flaw, a fracture mechanicsevaluation was performed that included a fatigue crack growth analysis to demonstratethat a small initial assumed flaw (10 percent through-wall), corresponding to theacceptance standards of ASME Code, Section Xl, Subarticle IWB-3500, would not growto quarter thickness during plant life. There are two potential time-dependencies in theCode Case N-481 evaluation:1. The fracture toughness of the cast austenitic stainless steel is nottime-dependent as the analysis used a lower bound fracture toughnessof 139-ksi'/in that bounds the saturated fracture toughness of theDavis-Besse material.2. The fatigue crack growth analysis is based on design cycles for a 40-yearplant life, and is therefore a TLAA requiring analysis and disposition forlicense renewal.The fatigue crack growth analysis assumed an initial flaw size corresponding to theacceptance standards of ASME Code Section Xl and considered all the significant planttransients. This analysis examined the design cycles and determined there were240 cycles that were significant to flaw growth in the RCPs. Then 2000 cycles wereconservatively analyzed, and flaw growth (initial 10 percent assumed through-wall hadgrown only to 15 percent through-wall) remained well below the quarter thicknesspostulated flaw. The analyzed cycles of 2000 bound the 60-year projected cycles shownin LRA Table 4.3-1, and the fatigue crack growth TLAA associated with the ASME Code
 
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5L-1 1-238Page 7 of 10Case N-481 evaluation will therefore remain valid for the period of extended operation inaccordance with 10 CFR 54.21(c)(1)(i).See the Enclosure to this letter for the revision to the DBNPS LRA.Question RAI XI.S8-1 Supplemental Response -Nuclear Safety-Related ProtectiveCoatings ProgramThe NRC initiated a telephone conference call with FENOC on July 27, 2011, todiscuss the FENOC response to RAI XI.S8-1 (submitted in FENOC letter datedJune 17, 2011 (ML11172A389)). The NRC staff asked for clarification as to whyASTM D5163 specifies a year-of-issue designator in part of the submittal but isnot consistent throughout the response.RESPONSE RAI XI.$8-1 Supplemental Response -Nuclear Safety-Related ProtectiveCoatings ProgramLRA Section B.2.42, "Nuclear Safety-Related Protective Coatings Program," is revisedto include the year-of-issue designator for (e.g., -91) to ASTM International (ASTM)standard numbers that did not include year-of-issue designators.See the Enclosure to this letter for the revision to the DBNPS LRA.Question Supplemental Response -Abandoned EquipmentThe NRC initiated a telephone conference call with FENOC on July 27, 2011, todiscuss the FENOC response to RAI 2.1-3 (submitted in FENOC letter datedApril 29, 2011 (ML11126A016)). The NRC staff was unclear from the FENOCresponse to RAI 2.1-3 if FENOC is aware of all the abandoned equipment. TheNRC staff stated that the current FENOC response does not meet the NRC'sexpectations and is not consistent with other applications. The NRC staffdescribed some methods that could meet the NRC's expectations:1. Identify all abandoned potentially fluid-filled equipment and verify that itis drained;2. Place all of the abandoned potentially fluid-filled equipment in scope andsubject it to aging management review; or,
 
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5L-1 1-238Page 8 of 103. Combinations of the two methods above.RESPONSE Supplemental Response -Abandoned EquipmentLRA Table A-1, "Davis-Besse License Renewal Commitments," Commitment 26, isrevised to ensure that abandoned equipment is identified, and either isolated anddrained or included within the scope of license renewal and subject to agingmanagement review prior to receipt of the renewed license.See the Enclosure to this letter for the revision to the DBNPS LRA.Question Supplemental Response -Methods for One Time InspectionsThe NRC initiated a telephone conference call with FENOC on July 27, 2011, todiscuss FENOC's response to RAI 3.3.2.2.4.3-1 and corresponding LRA SectionB.2.30 amendment in letter dated June 3, 2011 (ML11159A132). The amendment toLRA Section B.2.30 states that the "scope" program element is to include visualand volumetric inspections of the stainless steel makeup pump casings forcracking due to cyclic loading, but it does not state what type of visualexaminations will be used to detect cracking. The GALL [NUREG-1801] AMPXI.M32, "One-Time Inspection" states in the "detection of aging effects" programelement that the program manages cracking due to cyclic loading usingenhanced visual (EVT-1 or equivalent), surface, or volumetric examinations.However, some types of visual examination may not be sufficient to identifycracking, and it is unclear what visual examinations will be performed to meetthis need. The NRC requests the type of visual examination that will be used toidentify cracking as part of the One Time Inspection Program.RESPONSE Supplemental Response -Methods for One Time InspectionsLRA Sections A.1.30 and B.2.30, both titled "One-Time Inspection," are revised toinclude a table that identifies the type of inspection that will be used for detection ofaging effects.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
==Attachment==
5L-1 1-238Page 9 of 10Question Supplemental Response -OIN-352 -External Surfaces MonitoringProgramDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of April 25, 2011, NRC Inspectors requested that theDavis-Besse External Surfaces Monitoring Program document include a numberof enhancements. The FENOC License Renewal Project created Open ItemNumber (OIN)-352 to track the request, listed as follows:" Accessible components that credit the ESM Program for agingmanagement shall be inspected at least once per fuel cycle* Add acceptance criteria to the System Walkdown Check List" Add inspection parameters to the System Walkdown Check List, asfollows:-corrosion and material wastage (loss of material)-leakage from or onto external surfaces-worn, flaking, or oxide-coated surfaces-corrosion stains on thermal insulation-protective coating degradation (cracking and flaking)" Add a record retention requirement to retain the System Walkdown CheckList to document the results of the inspectionLRA Sections affected are A.1.15, B.2.15 and LRA Table A-1, "Davis-BesseLicense Renewal Commitments," Commitment 8 (enhancement commitment forthe External Surfaces Monitoring Program).RESPONSE Supplemental Response -OIN-352 -External Surfaces MonitoringProgramLRA Sections A.1.15 and B.2.15, both titled "External Surfaces Monitoring Program,"and LRA Table A-I, "Davis-Besse License Renewal Commitments," Commitment 8, arerevised to include the items identified above.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
==Attachment==
5L-1 1-238Page 10 of 10Question Supplemental Response -OIN-368 Fuel Oil ChemistryDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of May 9, 2011, NRC Inspectors requested that FENOCrevise the Fuel Oil Chemistry Program description to state that the programmonitors and trends water and particulate contamination concentrations inaccordance with the plant's technical specifications. Also, include anenhancement to the program to monitor and trend biological activity quarterly.RESPONSE Supplemental Response -OIN-368 Fuel Oil ChemistryLRA Sections A.1.20 and B.2.20, both titled "Fuel Oil Chemistry Program," and LRATable A-I, "Davis-Besse License Renewal Commitments," Commitment 28, are revisedto address the identified items.See the Enclosure to this letter for the revision to the DBNPS LRA.Question Supplemental Response -OIN-369 Reactor Vessel upper head lift lugsDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of May 9, 2011, NRC Inspectors identified that,contrary to the information provided in LRA Section 2.3.1.1, the reactor pressurevessel upper head lifting lugs have a "support" intended function relative to aheavy lift. Therefore, the lifting lugs are subject to aging management review, anda change to the LRA is required.RESPONSE Supplemental Response -OIN-369 Reactor Vessel upper head lift lugsFENOC agrees that the reactor pressure vessel (RPV) upper head lifting lugs have a"support" intended function relative to a heavy lift over safety related equipment.Therefore, the lifting lugs are subject to aging management review. LRA Section2.3.1.1, "Reactor Pressure Vessel," is revised to show that the RPV upper head liftinglugs are subject to aging management review. Also, LRA Table 2.3.1-1, "ReactorPressure Vessel Components Subject to Aging Management Review," is revised toinclude upper head lifting lugs with an intended function of support. In addition, a LRATable 3.1.2-1, "Aging Management Review Results -Reactor Pressure Vessel," isrevised to include the aging management review results of the upper head lifting lugs.See the Enclosure to this letter for the revision to the DBNPS LRA.
 
==Enclosure==
Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS)Letter L-11-238Amendment No. 13 to theDBNPS License Renewal ApplicationPage 1 of 98License Renewal ApplicationSections AffectedLRA Table of ContentsSection 2Section 2.3.1.1Table 2.3.1-1Table 2.4-12Section 3Table 3.1.2-1Table 3.1.2-4Section 3.2.2.1.3Section 3.2.2.1.4Section 3.2.2.2.3.6Table 3.2.1Table 3.2.2-2Table 3.2.2-3Table 3.2.2-4Table 3.2.2 P-S NotesSection 3.3.2.1.4Section 3.3.2.1.5Section 3.3.2.1.6Section 3.3.2.1.7Section 3.3.2.1.11Table 3.3.1Table 3.3.2-4Table 3.3.2-5Table 3.3.2-6Table 3.3.2-7Table 3.3.2-11Table 3.3.2-14Table 3.3.2-21Table 3.3.2-31Table 3.5.2-12Table 3.5.2 P-S NotesSection 4Table 4.1-1Section 4.3.4.2Table 4.3-2Section 4.6.1Section 4.7.6Section 4.8Appendix AAppendix A TOCSection A.1Section A. 1.8Section A. 1.9Section A. 1.15Section A. 1.20Section A. 1.22Section A.1.25Section A. 1.28Section A. 1.41Section A.2.5.1Section A.2.7.5Section A.2.8Table A-1Appendix BTable B-2Section B.2.8Section B.2.9Section B.2.15Section B.2.20Section B.2.22Section B.2.25Section B.2.28Section B.2.30Section B.2.41Section B.2.42The Enclosure identifies the change to the License Renewal Application (LRA) byAffected LRA Section, LRA Page No., and Affected Paragraph and Sentence. Thecount for the affected paragraph, sentence, bullet, etc. starts at the beginning of theaffected Section or at the top of the affected page, as appropriate. Below each sectionthe reason for the change is identified, and the sentence affected is printed in italics withdeleted text #,,-e4u and added text underlined.
 
==Enclosure==
L-1 1-238Page 2 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence2.3.1.1, and Pages 2.3-7 Reactor Vessel ExternalTable 2.3.1-1 thru 2.3-9 Attachments;Components Subject to AMR; and,New Table RowDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of May 9, 2011, NRC Inspectors identified that,contrary to the information provided in LRA Section 2.3.1.1, the reactor pressurevessel upper head lifting lugs have a "support" intended function relative to aheavy lift (OIN-369). LRA Section 2.3.1.1, "Reactor Pressure Vessel," and Table2.3.1-1, "Reactor Pressure Vessel Components Subject to Aging ManagementReview," are revised to read as follows:LRA Page 2.3-7:Reactor Vessel External AttachmentsThere are multiple external attachments to the reactor pressure vessel, includingthe teop-uyper head lifting lugs, insulation support pads, vessel handling lugs, andthe CRDM support skirt.LRA Page 2.3-8:Components Subject to AMRIn addition to those components specifically excluded in 10 CFR 54.21(a)(1)(i),such as instruments, the following components of the reactor pressure vessel arein the scope of license renewal, but are not subject to AMR:* O-rings and gasketsTop Head Liftinlg Lugs* Vessel Insulation Support Pads* Vessel Handling LugsThe internal attachments provide support to their respective components and allof the internal attachments are subject to AMR. External attachments are subjectto AMR if they are load bearing attachments connected to pressure retaining
 
==Enclosure==
L-1 1-238Page 3 of 98portions of the vessel with the exception of upper head lifting luqs. The upperhead lifting luogs have a "support" intended function relative to a heavy lift oversafety-related equipment and therefore, are subiect to AMR. The J-"iRg lugs.insulation support pads-, and vessel handling lugs do not bear significant weightduring power operation and are not subject to AMR. In addition, o-rings andgaskets are not designed for the life of the plant and are periodically replaced.LRA Page 2.3-9:Table 2.3.1-1Reactor Pressure VesselComponents Subject to Aging Management ReviewComponent Type Intended Function(as defined in Table 2.0-1)Upper head fiftinq lugqs Support
 
==Enclosure==
L-1 1-238Page 4 of 98Affected LRA SectionLRA Page No.Page 2.4-46Affected Paragraph and SentenceEDG Fuel Oil Storage Tank HoldDown Restraints rowTable 2.4-12In response to RAI 3.5.2.3.12-3, LRA Table 2.4-12, "Yard StructuresComponents Subject to Aging Management Review," is revised to delete the rowfor the emergency diesel generator fuel oil storage tank hold down restraints, toread as follows:Table 2.4-12Yard StructuresComponents Subject to Aging Management ReviewComponent Type Intended Function[(as defined in Table 2.0-1)-EDQG Fuej Q#! Strage T-ank HId- Down4 P Wa~tra [ SS.
 
==Enclosure==
L-1 1-238Page 5 of 98Affected LRA SectionLRA Page No.Page 3.1-59Affected ParaoraDh and SentenceTable 3.1.2-1New Row 100During the NRC Region III Inspection Procedure (IP) 71002, "License Renewal Inspection," held the week ofMay 9, 2011, NRC Inspectors identified that, contrary to the information provided in LRA Section 2.3.1.1, the reactorpressure vessel upper head lifting lugs have a "support" intended function relative to a heavy lift (OIN-369). LRATable 3.1.2-1, "Aging Management Review Results -Reactor Pressure Vessel," is revised to add a new row(Row No. 100) to address aging management of the upper head lifting lugs, and reads as follows:Table 3.1.2-1 Aging Management Review Results -Reactor Pressure VesselAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemAir with100 Upper Hea Suport Steel borated water Loss of Boric Acid Corrosion IV.A2-13 3.1.1-58 ALifting Lugs leaka_ e material(External)
 
==Enclosure==
L-1 1-238Page 6 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceTable 3.1.2-4 Page 3.1-169 Rows 30 and 31The NRC initiated a telephone conference call with FENOC on July 12, 2011, to discuss primary side fouling ofsteam generator tubes. In this Supplemental Response to the teleconference request, Rows 30 and 31 of LRATable 3.1.2-4, "Aging Management Review Results -Steam Generators," are revised to show as "Not used." LRATable 3.1.2-4 now reads as follows:
 
==Enclosure==
L-11-238Page 7 of 98Affected LRA Section LRA Page No. Affected Para-graph and Sentence3.2.2.1.3 Page 3.2-5 "Aging Management Programs"subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs" subsectionof LRA Section 3.2.2.1.3, "Core Flooding System," is revised to read as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Core Flooding System:* Bolting Integrity Program" Boric Acid Corrosion Program" External Surfaces Monitoring Program" Inspection of Internal Surfaces of Miscellaneous Pipinq and DuctingPro.gram* One-Time Inspection" PWR Water Chemistry Program
 
==Enclosure==
L-1 1-238Page 8 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.2.2.1.4 Page 3.2-6 "Aging Management Programs"subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs" subsectionof LRA Section 3.2.2.1.4, "Decay Heat Removal and Low Pressure InjectionSystem," is revised to read as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Decay Heat Removal and Low Pressure InjectionSystem:* Aboveground Steel Tanks Inspection* Bolting Integrity Program* Boric Acid Corrosion Program* Closed Cooling Water Chemistry Program* External Surfaces Monitoring Program* Inspection of Internal Surfaces of Miscellaneous Piping and DuctingProgram* Lubricating Oil Analysis Program* One-Time Inspection* PWR Water Chemistry Program* Selective Leaching Inspection
 
==Enclosure==
L-1 1-238Page 9 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.2.2.2.3.6 Page 3.2-9 Entire sectionIn response to RAI 3.2.2.2.3.6-2, Section 3.2.2.2.3.6, "Stainless Steel Piping,Piping Components, Piping Elements, and Tanks -Internal Condensation,"previously revised in FENOC Letter dated May 24, 2011 (ML1 1151A090), isrevised to read as follows:3.2.2.2.3.6 Stainless Steel Piping, Piping Components, Piping Elements,and Tanks -Internal CondensationLoss of material from pitting and crevice corrosion could occur for stainless steelpiping, piping components, piping elements, and tanks exposed to internalcondensation. Moist air is enveloped by the NUREG-1801 Chapter IX definitionof condensation. At Davis-Besse, loss of material at akr water intefaces forstainless steel piping, piping components, piping elements, and tanks that areexposed internally to moist air will be detected aRnd chaFr4terized by the OneTime Insection. Losas oPf matedial wh4ere contain~ants m~ay be concentrated bfrequent wetting and drying for stainless steel piping, pipng oponent&, tubing,and valve bies that a÷re exposed internal. y t .o.. ar. hich is e.. ..cped bythe ^UPREG 1801 chapter IX definition Of will be managed by theplant-specific Inspection of Internal Surfaces in Miscellaneous Piping and DuctingProgram.
 
==Enclosure==
L-1 1-238Page 10 of 98Affected LRA SectionLRA Page No.Affected Para-graph and SentenceTable 3.2.1Page 3.2-17Row 3.2.1-08 "Discussion"Text in "Discussion" column is revised based on the response to RAI 3.2.2.2.3.6-2. LRA Table 3.2.1, "Summary ofAging Management Programs for Engineered Safety Systems Evaluated in Chapter VII of NUREG-1801," and nowreads as follows:Table 3.2.1 Summary of Aging Management Programs for Engineered Safety Features SystemsEvaluated in Chapter V of NUREG-1801Item Aging Effect/ Aging Management Evaluation DiscussionNumber Component/Commodity Mechanism Programs Recommended3.2.1-08 Stainless steel piping, piping Loss of material due A plant-specific aging Yes, plant- Consistent with NUREG-1 801.components, piping elements, to pitting and crevice management program is specific Loss of material due to pittingand tank internal surfaces corrosion to be evaluated, and crevice corrosion at-aii-wateexposed to condensation inte.t. -a.oce in stainless steel(internal) piping, piping components,piping elements, and tanks thatare exposed to moist air(internal) will b-e d-t-tled and.hrcozewmd by the One Timeplnt-pectifie nsetinoInternal Surcmay beGencontrated by freguent wettingand drying for stainless steel.iig pipi.ng cOm~ponentsoexposed internally to moeist iO-ntemnay9will be managed by theplant-specific Inspection ofInternal Surfaces in
 
==Enclosure==
L-1 1-238Page 11 of 98Table 3.2.1 Summary of Aging Management Programs for Engineered Safety Features SystemsEvaluated in Chapter V of NUREG-1 801Item Aging Effect/ Aging Management FurtherNumber Component/Commodity Mechanism Programs EcommendedRecommendedMiscellaneous Piping andDucting Program.Further evaluation isdocumented in Section3.2.2.2.3.6.
 
==Enclosure==
L-1 1-238Page 12 of 98Affected LRA SectionLRA Page No.Page 3.2-53Affected Paragraph and SentenceTable 3.2.2-2Row 21In response to RAI 3.2.2.2.3.6-2, row 21 of LRA Table 3.2.2-2, "Aging Management Review Results -ContainmentSpray System," is revised as follows:Table 3.2.2-2 Aging Management Review Results -Containment Spray SystemNUREG-Row Component Intended Aging Effect Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemOne Timno Inptenal EPressure Stainless Moist air Loss of Inspection of InteVnal E21 Piping boundary Steel (Internal) material Surfaces in V.A-26 3.2.1-08 022Miscellaneous Pipinq 0210and Ducting
 
==Enclosure==
L-1 1-238Page 13 of 98Affected LRA SectionLRA Page No.Page 3.2-63Affected Paragraph and SentenceTable 3.2.2.3Row 25In response to RAI 3.2.2.2.3.6-2, row 25 of LRA Table 3.2.2-3, "Aging Management Review Results -CoreFlooding System," is revised as follows:Table 3.2.2-3 Aging Management Review Results -Core Flooding SystemNUREG-Row Component Intended Aging Effect Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemneTime !nRSP9cGPressure Stainless Moist air Loss of Inspection of Internal E25 Piping Pressure Stanle MoIternair LossriofSurfaces in V.D1 -29 3.2.1-08 0202boundary Steel (Internal) material Miscellaneous PipinQ 0210and Ductinq
 
==Enclosure==
L-1 1-238Page 14of 98Affected LRA SectionLRA Page No.Pages 3.2-86 &3.2-87Affected Para-graDh and SentenceTable 3.2.2-4Rows 113 & 119In response to RAI 3.2.2.2.3.6-2, rows 113 and 119 of LRA Table 3.2.2-4, "Aging Management Review Results -Decay Heat Removal and Low Pressure Injection System," is revised as follows:Table 3.2.2-4 Aging Management Review Results -Decay Heat Removal and Low Pressure Injection Systemg Effet NUREG-Row Component Intended Material ing Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItem,neThime olfspectienTank -BWST Pressure Stainless Moist air Loss of Inspection of Internal E113 Surfaces i V.01-29 3.2.1-08 0210(DB-T10) boundary Steel (Internal) material Surfaces inMiscellaneous Pipina 02i4and DuctinaOne-Time InspecioTank -Incore Pressure Stainless Moist air Loss of Inspection of Internal E119 instrument Pressure Stail MIsteair LossroflSurfaces in V.D1-29 3.2.1-08 0210tank (DB-T92) boundary Steel (internal) material Miscellaneous Pipina 02144and Ductinq
 
==Enclosure==
L-1 1-238Page 15 of 98Affected LRA SectionSection 3.2Plant-Specific NotesLRA Page No.Page 3.2-118Affected Para-graph and SentenceRow 0211In response to RAI 3.2.2.2.3.6-2, row 0211 of Section 3.2, "Plant-Specific Notes," is no longer used, and is revisedas follows:Plant-Specific Notes:0211 The One T;me Inspection w... confrfm, for components t to "Maoist air (Internal)".envonment at the air- water int,.;-e, theNabt sedne of aging effoct6 or that aging is 6low aGtng so as to not affect the subject component's intendod fyntion duing the peof extended oper-ation, -which mvelfios the effeortivonP eas of aging managemgent proegram6 credited above and beiow this into ace.Not used
 
==Enclosure==
L-1 1-238Page 16 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.3.2.1.4 Page 3.3-9 "Aging Management Programs"subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs" subsectionof LRA Section 3.3.2.1.4, "Boron Recovery System," is revised to read as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Boron Recovery System:* Bolting Integrity Program* Closed Cooling Water Chemistry Program" Inspection of Internal Surfaces of Miscellaneous Piping and DuctingPro.gram* One-Time Inspection" PWR Water Chemistry Program
 
==Enclosure==
L-1 1-238Page 17 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.3.2.1.5 Page 3.3-10 "Aging Management Programs"subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs"subsection of LRA Section 3.3.2.1.5, "Chemical Addition System," is revised toread as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Chemical Addition System:* Bolting Integrity Program* Inspection of Internal Surfaces of Miscellaneous Piping and DuctingProgram* One-Time Inspection* PWR Water Chemistry Program
 
==Enclosure==
L-1 1-238Page 18 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.3.2.1.6 Page 3.3-11 "Aging Management Programs"subsectionIn response to RAI 3.3.2.2.5-2, and to correct the inadvertent assignment of theFlow-Accelerated Corrosion (FAC) Program in FENOC Letter dated May 24, 2011(ML11151A090), the "Aging Management Programs" subsection of LRASection 3.3.2.1.6, "Circulating Water System," is revised to read as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Circulating Water System:" Bolting Integrity Program" Collection, Drainage, and Treatment Components Inspection Pro-gram" External Surfaces Monitoring ProgramF49 AcoIerated Corrosion (FAGC) Program* Inspection of Internal Surfaces of Miscellaneous Piping and DuctingProgram* Open-Cycle Cooling Water Program
 
==Enclosure==
L-1 1-238Page 19 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence3.3.2.1.7 Page 3.3-12 "Aging Management Programs"subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs" subsectionof LRA Section 3.3.2.1.7, "Component Cooling Water System," is revised to readas follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Component Cooling Water System:" Bolting Integrity Program" Boric Acid Corrosion Program* Closed Cooling Water Chemistry Program* External Surfaces Monitoring Program* Inspection of Internal Surfaces of Miscellaneous Piping and DuctinQPro-gram* One-Time Inspection" Open-Cycle Cooling Water Program
 
==Enclosure==
L-1 1-238Page 20 of 98Affected LRA Section LRA Page No. Affected Para-graDh and Sentence3.3.2.1.11 Page 3.3-15 & "Aging Management Programs"3.3-16 subsectionIn response to RAI 3.2.2.2.3.6-2, the "Aging Management Programs" subsectionof LRA Section 3.3.2.1.11, "Demineralized Water Storage System," is revised toread as follows:Aging Management ProgramsThe following aging management programs manage the aging effects for subjectmechanical components of the Demineralized Water Storage System:" Bolting Integrity Program* Boric Acid Corrosion Program* External Surfaces Monitoring Program* Inspection of Internal Surfaces of Miscellaneous Piping and DuctinqPro-gram* One-Time Inspection* PWR Water Chemistry Program
 
==Enclosure==
L-1 1-238Page 21 of 98Affected LRA SectionLRA Page No.Pages 3.3-80Affected Paragraph and SentenceRow 3.3.1-49, "Discussion" columnTable 3.3.1Text in "Discussion" column is revised based on the response to RAI 3.3.1.49-2. LRA Table 3.3.1, "Summary of AgingManagement Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1 801," now reads as follows:Table 3.3.1 Summary of Aging Management Programs for Auxiliary SystemsEvaluated in Chapter VII of NUREG-1801Item Aging Effect/ Aging Management FurtherIter Component/Commodity Mechanism Programs Evaluation DiscussionNumber Recommended3.3.1-49 Stainless steel; steel with Loss of material Closed-Cycle Cooling No Net appfeable7stainless steel cladding heat due to Water System Based on plant-specific operatingexchanger components microbiologically experience, Mloss of material dueexposed to closed cycle cooling influenced to microbiologically influencedwater corrosiontomcoilgclynfucecorrosion is not identified as anaging effect requiringmanagement for stainless steelheat exchanger components thatare exposed to closed cyclecooling water. However, theexisting Closed Cooling WaterChemistry Program requires thatsystems within the scope of theprogram are monitored for thepresence of microbiologicalactivity in accordance with theEPRI Closed-Cycle CoolingWater guidelines.In addition, there are no steelwith stainless steel cladding heatexchanger components that are
 
==Enclosure==
L-1 1-238Page 22 of 98Table 3.3.1 Summary of Aging Management Programs for Auxiliary SystemsEvaluated in Chapter VII of NUREG-1801Item Aging Effect/ Aging Management FurtherNumber Component/Commodity Aginfec tManiragmet Evaluation DiscussionNumber mMechanism Programs Recommendedexposed to closed cycle coolingwater and subject to agingmanagement review.
 
==Enclosure==
L-1 1-238Page 23 of 98Affected LRA SectionLRA Page No.Pages 3.3-98Affected Para-graph and SentenceRow 3.3.1-75, "Discussion" columnTable 3.3.1Text in "Discussion" column is revised based on the response to RAI 3.3.2.2.5-2. LRA Table 3.3.1, "Summaryof Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801," now readsas follows:Table 3.3.1 Summary of Aging Management Programs for Auxiliary SystemsEvaluated in Chapter VII of NUREG-1801FurtherItem Aging Effect/ Aging Management Euatis sNumer Component/Commodity Mehns rgasEvaluation DiscussionNumber oMechanism Programs Recommended3.3.1-75 Elastomer seals and Hardening and loss Open-Cycle Cooling No Consistent with NUREG-1 801,components exposed to raw of strength due to Water System but a different agingwater elastomer management program isdegradation; loss assigned.of material due to Hardening and loss of strengtherosion Hreigadls fsrntfor elastomer components thatare exposed to raw water will bemanaged by the A6P9GUen-e,iin 9 iPoeg~am Collection,Drainage, and TreatmentComponents InspectionProgram.
 
==Enclosure==
L-1 1-238Page 24 of 98Affected LRA SectionLRA Page No.Page 3.3-211Affected Paragraph and SentenceTable 3.3.2-4Row 158In response to RAI 3.2.2.2.3.6-2, row 158 of LRA Table 3.3.2-4, "Aging Management Review Results -BoronRecovery System," is revised as follows:Table 3.3.2-4 Aging Management Review Results -Boron Recovery SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemTank -One Tim e .n. pe, ti. .Concentrates Structural Stainless Moist air Loss of Inspection of Internal Estorage tank integrity Steel (Internal) material Scellaneous 033(DBT1 6) Miscellaneous 03321(1B-T116) Pipingq and Ductingq
 
==Enclosure==
L-1 1-238Page 25 of 98Affected LRA SectionLRA Page No.Page 3.3-225Affected Paragraph and SentenceTable 3.3.2-5Row 59In response to RAI 3.2.2.2.3.6-2, row 59 of LRA Table 3.3.2-5, "Aging Management Review Results -ChemicalAddition System," is revised as follows:Table 3.3.2-5 Aging Management Review Results -Chemical Addition SystemdAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I NoteNo. Type Function(s) Management Program Volume 2 ItemItemTank -Boric ...Ai. .. "..P.c ...acid addition Pressure Stainless Moist air Loss of Inspection of Internal Etanks (DB- boundary Steel (Internal) material Sceslaneous 033T71 &2) Miscellaneous 0332T1-1 1 2)11 Pipingq and Ductingq
 
==Enclosure==
L-1 1-238Page 26 of 98Affected LRA SectionLRA Paae No.Page 3.3-231Affected Paraaraph and SentenceTable 3.3.2-6Row 4 and 1 [New] RowIn response to RAI 3.3.2.2.5-2, row 4 and a previously added "raw water" row (see FENOC Letter datedJune 3, 2011 (ML 1 159A1 32)) of LRA Table 3.3.2-6, "Aging Management Review Results -Circulating WaterSystem," are revised as follows:Table 3.3.2-6 Aging Management Review Results -Circulating Water SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemlnSpoction of InternalHardening and Df..c ...4tFlexible Structural Raw water loss of ""and D.. .ti. VII.C1-1 3.3.1-75 EConnection integrity (Internal) strength TCollection, inanestrengthand TreatmentComponentsInspectionlAR;P90in of IntarlnaFlexible Structural Elastomer Raw water Loss of Poing and n VII.C-2 3.3.1-75 EConnection integrity (Internal) material Collection, Drainage,and TreatmentComponentsInspection
 
==Enclosure==
L-1 1-238Page 27 of 98Affected LRA SectionLRA Page No.Page 3.3-246Affected Paragraph and SentenceTable 3.3.2-7Row 80In response to RAI 3.2.2.2.3.6-2, row 80 of LRA Table 3.3.2-7, "Aging Management Review Results -ComponentCooling Water System," is revised as follows:Table 3.3.2-7 Aging Management Review Results -Component Cooling Water SystemNUREG-Ro Cmonn ItnddAging Effect Aging Management 1801, Table IRow Component Intended Material Environment Requiring Agn aae et 10, Tbe1 NotesNo. Type Function(s) Management Program Volume 2 ItemItemT a n k -n sp i e c ti o n o In te rn al EChemical pot Structural Stainless Moist air Loss of Inspection of InteVnal E80 feeder (DB- integrity Steel (Internal) material Surfaces in V.01-29 3.2.1-08 034Ti Miscellaneous 03321 PipinQ and Ductingq
 
==Enclosure==
L-1 1-238Page 28 of 98Affected LRA SectionLRA Page No.Page 3.3-276Affected ParaaraDh and SentenceTable 3.3.2-11Row 29In response to RAI 3.2.2.2.3.6-2, row 29 of LRA Table 3.3.2-11, "Aging Management Review Results -Demineralized Water Storage System," is revised as follows:Table 3.3.2-11 Aging Management Review Results -Demineralized Water Storage SystemNUREG-Aging Effect nNotesRow Component Intended Material Environment Requiring1801, TableNo. Type Function(s) Management Program Volume 2 ItemItemTank- Lab One Tinge "ngpe....ndemin. water Structural Stainless Moist air Loss of Inspection of Inter.al Estorage tank integrity Steel (Internal) material Scellaneous 033(DB-T1 08) Miscellaneous 03321 (1B1T108) Piping and Ducting
 
==Enclosure==
L-1 1-238Page 29 of 98Affected LRA SectionLRA Page No.Page 3.3-334Affected Paraaralh and SentenceTable 3.3.2-14Row 172 and I [New] RowIn response to RAI 3.3.2.2.5-2, row 172 and a previously added "raw water" row (see FENOC Letter datedJune 3, 2011 (MI111 59A132)) of LRA Table 3.3.2-14, "Aging Management Review Results -Fire ProtectionSystem," are revised as follows:Table 3.3.2-14 Aging Management Review Results -Fire Protection SystemNUREG-Ro Cmonn ItnddAging Effect Aging Management 1801, Table 1Row Component Intended Material Environment Requiring Agn a ae et 10 , T be1 NotesNo. Type Function(s) Management Program Volume 2 ItemItemAn~peectien of lfater-n3!F le x ib le P r e s s u r e R a w w a t e r H a r d e n in g a n d Pi p ng ...n De ', , n ,172 Elastomer loss of '0i"'9 ....... "` VII.C1-1 3.3.1-75EConnection boundary RawIwaternalosstrofgCollection, Drainage,y(Internal) strength and TreatmentComponentsInspectioninspeotien of Intorn-alAMiaGe#aneousFlexible Pressure Raw water Loss of Ping and , VII.C1-2 3.3.1-75 EConnection boundary (Internal) material Collection, Drainage,and TreatmentComponentsLnspection
 
==Enclosure==
L-1 1-238Page 30 of 98Affected LRA SectionLRA Pane No.Page 3.3-405Affected Para-graDh and SentenceTable 3.3.2-21Row 16 and I [New] RowIn response to RAI 3.3.2.2.5-2, row 16 and a previously added "raw water" row (see FENOC Letter datedJune 3, 2011 (ML1 1159A1 32)) of LRA Table 3.3.2-21, "Aging Management Review Results -Miscellaneous LiquidRadwaste System," are revised as follows:Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemlnspotion of Internsl16 Flexible Structural ElwstomerHardening and M 9 .. .. tin,nectio itgrita Elastomer water loss of " "" 4...... .VII.C1-1 3.3.1-75 EConnection integrity (Internal) strength Collection, Drainage,and TreatmentComponentsInspectionof Interne,Mi8G9#an~eu6Flexible Structural Elastomer Raw water Loss of ,-ng ... "tng VII.C1-2 3.3.1-75 EConnection integrity (Internal) material Collection, Drainage.and TreatmentComponentsInspection
 
==Enclosure==
L-1 1-238Page 31 of 98Affected LRA SectionLRA Page No.Page 3.3-533Affected Para-graph and SentenceTable 3.3.2-31Row 20During the development of the response to RAI 3.2.2.2.3.6-2, an inadvertent omission was identified in FENOCletter dated May 24, 2011 (ML1 1151 A090). Specifically, in the Enclosure to the letter, page 110 of 206, row 20 ofLRA Table 3.3.2-31, "Aging Management Review Results -Station Plumbing, Drains, and Sumps System," plant-specific note 0332 should have been included in the revised row "Notes" column. Row 20 of LRA Table 3.3.2-31 isrevised to include plant-specific note 0332 as follows:Table 3.3.2-31 Aging Management Review Results -Station Plumbing, Drains, and Sumps SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 NoteNo. Type Function(s) MalagenengProgram Volume 2 ItemManagement ItemInspection of Internal E20 Piping Pressure Stainless Moist air Loss of Surfaces in V.01-29 3.2.1-08 0332boundary Steel (Internal) material Miscellaneous 0334Piping and Ducting 0334
 
==Enclosure==
L-1 1-238Page 32 of 98Affected LRA SectionLRA Pane No.Page 3.5-112Affected Paragraph and SentenceTable 3.5.2-12 andPlant-specific NotesRows 7 and 8, andNote 0531In response to RAI 3.5.2.3.12-3, LRA Table 3.5.2-12, "Aging Management Review Results -Yard Structures," isrevised to delete rows 7 and 8 for the emergency diesel generator fuel oil storage tank hold down restraints. Also,as part of this revision, Plant-specific Note 0531 is no longer used, and is deleted. LRA Table 3.5.2-12 andTable 3.5.2 Plant-Specific Notes are revised to read as follows:Table 3.5.2-12 Aging Management Review Results -Yard StructuresNUREG-Row Component Intended Aging Effect Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 ItemItemSte~age Tank rh7 Ho-ldown .SR Rtan N/ene Anl=J -24 .3.-9 CRes-aiksNot usedEeL-Oilst -Qeb tr n w ti intoi irai8 Upied-Down SSR VaNeaGAne MegPIeFiRn 053iRee~tsaNot usedPlant-Specific Notes:0531 NUREG 1801 does not l... a " t""cturi bkfile onvironment for q tsite9 Noagig9.fft8requin mn."nt Wuseidnntdifid fe-F the EDG Fuel Oi# Storage Tank hold don wie rpei a StrUctural bacrkfil environment. Howe er, the identifid AMP14i1l beh69o to con firmn the ab hse nce of signifiant gn ffetsfo tho period of extended operation. The Str-wct-r-al bacyrkFil is aboevegrade and the elovation locr-ation -of theworpeiabethst'sgunaerlvto. Not used
 
==Enclosure==
L-11-238Page 33 of 98Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable 4.1-1Page 4.1-4New rowThe NRC initiated a telephone conference call with FENOC on July 15, 2011, todiscuss the FENOC response to RAI 4.1-2 (ML11 172A389). The NRC does notagree that the cycle-dependent fatigue flaw growth analysis is not a time-limitedaging analysis. FENOC provides this supplemental response to RAI 4.6-1 to adda new LRA Section 4.7.6, "ASME Code Case N-481 Evaluation." LRA Table4.1-1, "Time-Limited Aging Analyses," is revised to include a new row todisposition new LRA Section 4.7.6, as follows:Table 4.1 1 Time-Limited Aging AnalysesResults of TLAA Evaluation by Category 54.21 (c)(1) LRAParagraph SectionOther Plant-Specific Time-Limited Aging Analyses 4.7ASME Code Case N-481 Evaluation Li) 4.7.6
 
==Enclosure==
L-1 1-238Page 34 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence4.3.4.2 Page 4.3-28 Entire sectionTable 4.3-2 Page 4.3-31 Table values and footnotesThe NRC initiated a telephone conference call with FENOC on July 12, 2011, todiscuss the FENOC response to RAI 4.3-17 (ML11172A389) related to theenvironmentally-assisted fatigue (EAF) evaluation results for the surge linepiping. FENOC provides this supplemental response to RAI 4.3-17 to revise LRASection 4.3.4.2, "Davis-Besse Evaluation -Surge Line Fatigue Results," andLRA Table 4.3-2, "Davis-Besse CUFs for NUREG/CR-6260 Locations," whichnow read as follows:4.3.4.2 Davis-Besse EvaluationSurge Line Fatigue ResultsThe bounding environmentally adjusted cumulative usage factors for the surgeline are as follows:" The maximum design CUF for the stainless steel pipe adjacent to theoutboard end of the hot leg surge nozzle is 0.179. In-air CUF (adiusted CUF)based on 60-year proiected cycles (except for best estimate 60-year proiectcycles of 114 used for HU/CDs events) is 0.066. Using the integrated Fenapproach described above, the Uen for the stainless steel pipe adjacent to theoutboard end of the hot leg surge nozzle weld overlay is 0.387 with a globalFen of 5.83. An apfi-'sted CUF Gf 0.07, s ebtaigd b,3, ýd, dng th- Uof0.387" The maximum design CUF for the elbows is 0.643. In-air CUF (adiusted CUF)based on 60-year proiected cycles (except for best estimate 60-year proiectcycles of 114 used for HU/CDs events) is 0.239. Using the integrated Fenapproach described above, the maximum Uen for the elbows is 0.996 with aglobal Fen of 4.17. An adjusted , UF -of 0.:239 is obtained by d ng Me ,e,,, ,_ef0.0096 by the global f~e---" The maximum design CUF for the straight pipe is 0.764. In-air CUF (adjustedCUF) based on 60-year proiected cycles (except for best estimate 60-yearproject cycles of 114 used for HU/CDs events) is 0.336. Using the integratedFen approach described above, the maximum Uen for the straight pipe is 0.846with a global Fen of 2.52. An adjusted CUE of 0.336 is obtained by dividig theU.~o 0.46 y te goba frof-2.52.
 
==Enclosure==
L-1 1-238Page 35 of 989 The maximum design CUF for the stainless steel weld that connects thesurge line to the pressurizer surge nozzle safe end is 0.51. In-air CUF(adiusted CUF) based on 60-year proiected cycles (except for best estimate60-year proiect cycles of 114 used for HU/CDs events) is 0.073. Using theintegrated Fen approach described above, the Uen for the stainless steel weldthat connects the surge line to the pressurizer surge nozzle safe end is 0.644with a global Fen of 8.84. An adyusted ,UF of 0.07-3 i6 obtained by dvding th,U" of 0.644 by the glb-al & of.4.See the revision to LRA Table 4.3-2 on the next page.
 
==Enclosure==
L-1 1-238Page 36 of 98Table 4.3-2 Davis-Besse CUFs for NUREG/CR-6260 LocationsNUREG/CR-6260 Material Design Adjustedgeneric locations Davis-Besse plant-specific locations type CUFs CUFs Fen UenReactor vessel shell and lower Vessel shell and lower head LAS 0.024 NA8  2.45 0.059head Incore instrument nozzle NBA 0.770 0.2065 4.16 0.857Reactor vessel inlet and outlet Reactor vessel inlet nozzle LAS 0.829 0.1461 2.45 0.3582 nozzles Reactor vessel outlet nozzle LAS 0.768 0.3351 2.45 0.8213 Pressurizer surge line Hot leg surge nozzle inside radius CS 0.445 NA8  1.74 0.774Piping adjacent to outboard end of hot leg surge nozzle SS 0.179 0,066 5.83 0.387Piping elbows SS 0.643 0.2392 4.17- 0.996"Piping straights SS 0.764 0.3362 2.52- 0.846-Piping to pressurizer surge nozzle safe end weld, SS 0.51 0.0732 8.842 0.6442Pressurizer surge nozzle inside radius CS 0.182 NA8  1.74 0.317Pressurizer surge nozzle, safe end SS 0.108 0.0581 15.35 0.8924 HPI/Makeup nozzle HPI/Makeup nozzle CS 0.589 0.348' 1.74 0.606HPI/Makeup nozzle safe end SS 0.664 0.5504 8.036 4.41775 Reactor vessel core flood nozzle Nozzle LAS 0.0504 NA8  2.45 0.1236 Decay heat Class 1 piping Decay heat to core flood tee SS 0.233 NA82.55 0.5951. Adjusted CUF obtained by identifying incremental fatigue contribution attributed to the full NSSS design transient cycles for design CUF and reducingthose incremental contributions based on the 60-year cycle projections.L.3.4.5.6.7.8.9.the pr,....rizer. surg. line. Adiusted CUF was calculated usincq 60-year proiected cycles (except for best estimate 60-year proiect cycles of 114 used forHU/CDs events).Design CUF reduced from 0.589 to 0.348 by removing conservatisms in the original calculation. Full set of design cycles were used for the calculation.Design CUF reduced from 0.664 to 0.550 by removing conservatisms in the original calculation. Full set of design cycles were used for the calculation.Adjusted CUF obtained by applying the alternating stresses from the original design calculation to the new in-air design curve in NUREG/CR-6909 forstainless steel.This is a global Fen obtained by dividing Uen by the CUF (4.417/ 0.550).4.417 is >1.0 and is unacceptable for the period of extended operation. (See Section 4.3.4.2, Location 4).Adjusted CUF was not required. Design CUF multiplied by Fen resulted in an Uen of < 1.0.Fen was determined for each transient pair (integrated Fen approach), the Uep for each pair is determined by multiplying the in-air usage (adjusted CUF) forthat transient pair by the Fen calculated for that pair. The Uen for each transient pair were added to come up with cumulative Uen for that specific location.Fen presented in the Table is a global Fen calculated by dividing the cumulative U.n by the adjusted CUF.
 
==Enclosure==
L-1 1-238Page 37 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence4.6.1 Page 4.6-1 Second paragraphIn response to RAI 4.6-1, LRA Section 4.6.1, "Containment Vessel," secondparagraph, is revised to read as follows:4.6.1 CONTAINMENT VESSELThe containment vessel is a cylindrical steel pressure vessel with hemisphericaldome and ellipsoidal bottom which houses the reactor vessel, reactor coolantpiping, pressurizer, pressurizer quench tank and coolers, reactor coolant pumps,steam generators, core flooding tanks, letdown coolers, and normal ventilatingsystem. The containment vessel is a Class B vessel as defined in the ASMESection III, Paragraph N-132, 1968 Edition through Summer 1969 Addenda.The containment vessel is designed to resist dead loads, LOCA loads, operatingloads, external pressure load, temperature and pressure, impingement force andmissiles, wind loads, seismic loads, gravity loads, and live loads. Thecontainment vessel meets the requirements of ASME Section III, ParagraphN-415.1; thereby justifying the exclusion of cyclic or fatigue analyses in thedesign of the containment vessel. Analysis of 400 pressure cycles(from -0.67 psiq to 45 psIQ -25 to !20 psi) and 400 temperature cycles(from 30&deg;F to 1200F) were performed against the requirements of ASMESection III, Paragraph N-415.1. To date, the containm.nt vessel hag not soonany pressure cyles f-r25 to 1-:20 psi. The 60-year proiected cycles for plantheatup and cooldown are 128 (shown in Table 4.3-1) and are less than thespecified 400 pressure cycles and 400 temperature cycles. Therefore, the valuesof 400 pressure and temperature cycles used to exclude fatigue analyses will notbe exceeded for 60 years of operation. Thus, the TLAAs associated withexclusion of fatigue analyses for the containment vessel will remain valid for theperiod of extended operation.Disposition: 10 CFR 54.21(c)(1)(i) The TLAAs excluding the containmentvessel from fatigue analysis per ASMESection III, Paragraph N415-1 willremain valid through the period ofextended operation.
 
==Enclosure==
L-1 1-238Page 38 of 98Affected LRA Section LRA Page No. Affected Paragraph and Sentence4.7.6, Page 4.7-6 New Section4.8 Page 4.8-2 New ReferenceLRA Table of Page xii New entryContentsThe NRC initiated a telephone conference call with FENOC on July 15, 2011, todiscuss the FENOC response to RAI 4.1-2 (ML11172A389). The NRC does notagree that the cycle-dependent fatigue flaw growth analysis is not a time-limitedaging analysis. FENOC provides this supplemental response to RAI 4.6-1 to adda new LRA Section 4.7.6, "ASME Code Case N-481 Evaluation." LRA Section4.8 is revised to include a new reference to support new Section 4.7.6. The LRATable of Contents, not presented below, is revised to include new Section 4.7.6.New LRA Section 4.7.6 and the new reference in LRA Section 4.8 readas follows:476 A SIME CODE CA SE/-481/EVAL LIA TIONThe reactor coolant pumps (RCPs) are the only ASME Code Class I pumpsinstalled at Davis-Besse. The Pump casings are constructed of cast austeniticstainless steel. The applicable ASME Code for the current Third Ten-YearInspection Interval for Davis-Besse is ASME Section X1, 1995 Edition, throughthe 1996 Addenda, as modified by 10 CFR 50.55a or relief granted inaccordance with 10 CFR 50.55a. Examination Category B-L-1 of this Code yearrequires volumetric examination on pump casinq welds. ASME Code CaseN-481, "Alternative Examination Requirements for Cast Austenitic PumpCasings," provides an alternative to the volumetric examination requirement. Thiscode case allows the replacement of volumetric examinations of primary looppump casings with fracture mechanics-based integrity evaluation (Item (d) of thecode case) supplemented by specific visual examinations. Davis-Besse hasinvoked the use of Code Case N-481 in place of the volumetric examinationrequirements of Code Category B-L-1. The NRC has accepted Code Case N-481for use in inservice inspection programs.Code Case N-481 requires an evaluation to demonstrate the safety andserviceability of the pump casings. The evaluation for the Davis-Besse RCPsrequired by Code Case N-481 is documented in Structural Integrity Associates(SIA) report SIR-99-040 [Reference 4.8-181. This evaluation assumed a quarterthickness flaw, with length six times its depth, and showed that the flaw willremain stable considering the stresses and material properties of the pumpcasing. To determine stability of the postulated flaw, a fracture mechanics
 
==Enclosure==
L-1 1-238Page 39 of 98evaluation was performed that included a fatique crack growth analysis todemonstrate that a small initial assumed flaw (10 percent through-wall),corresponding to the acceptance standards of ASME Code, Section Xl,Subarticle IWB-3500, would not grow to quarter thickness during plant life.There are two potential time-dependencies in the Code Case N-481 evaluation.1. The fracture toughness of the cast austenitic stainless steel is not timedependent as the analysis used a lower bound fracture toughness of139 ksiin that bounds the saturated fracture toughness of theDavis-Besse materiaL2. The fatique crack growth analysis is based on desiqn cycles for a40 year plant life and therefore, is a TLAA requiring analysis anddisposition for license renewal.The fatique crack growth analysis assumed an initial flaw size corresponding tothe acceptance standards of ASME Code Section X1 and considered all thesignificant plant transients. This analysis examined the desiqn cycles anddetermined there were 240 cycles that were significant to flaw growth in theRCPs. Then 2000 cycles were conservatively analyzed, and flaw growth (initial10 percent assumed through-wall had grown only to 15 percent through-wall)remained well below the quarter thickness postulated flaw. The analyzed cyclesof 2000 bound the 60-year proiected cycles shown in LRA Table 4.3-1 andtherefore, the fatique crack growth TLAA associated with the ASME Code CaseN-481 evaluation will remain valid for the period of extended operation.Disposition: 10 CFR 54.21(c)(1)(i) The fatique crack growth TLAAassociated with ASME Code CaseN-481 evaluation will remain validthrough the period of extendedoperation.4.8 REFERENCES4.8-18 Structural Integrity Associates Report SIR-99-040, "ASME Code CaseN-481, Evaluation of Davis-Besse Reactor Coolant Pumps" Rev. 1,September 2000 (ADAMS Accession No. MLOI 1200090)
 
==Enclosure==
L-1 1-238Page 40 of 98Affected LRA Section LRA Page No. Affected Para-graph and SentenceA.1 Page A-9 New third paragraphThe NRC initiated a telephone conference call with FENOC on July 12, 2011, todiscuss the FENOC response to RAI B.1.4-1. FENOC provides thissupplemental response to RAI B.1.4-1 to add a new third paragraph toLRA Section A.1, "Summary Descriptions of Aging Management Programs andActivities," as follows:A.1 SUMMARY DESCRIPTIONS OF AGING MANAGEMENTPROGRAMS AND ACTIVITIESExisting FENOC processes require reviews of relevant site and industryoperating experience and periodic benchmarkinq to ensure programenhancements are identified and implemented. Such ongoing reviews identifypotential needs for a-ging management program revisions to ensure theireffectiveness throughout the period of extended operation.
 
==Enclosure==
L-1 1-238Page 41 of 98Affected LRA Section LRA Page No. Affected Para-graph and SentenceA.1.8 Page A-1I Second paragraph,new last sentenceIn response to RAI 3.3.1.49-1, the second paragraph of LRA Section A.1.8,"Closed Cooling Water Chemistry Program," previously revised in FENOC Letterdated May 24, 2011 (ML1 1 151A090), is revised to read as follows:A.1.8 CLOSED COOLING WATER CHEMISTRY PROGRAMAlso, the Closed Cooling Water Chemistry Program includes corrosion ratemeasurement at selected locations in the closed cooling water systems. Inaddition, periodic inspections of opportunity will be conducted when componentsare opened for maintenance, repair, or surveillance, to ensure that the existingenvironmental conditions are not causing material degradation that could result ina loss of component intended function during the period of extended operation.A representative sample of piping and components will be inspected on a10-year interval, with the first inspection taking place prior to entering the periodof extended operation. Systems within the scope of this program are monitoredfor the presence of microbioloqical activity in accordance with the EPRI Closed-Cycle Coolina Water auidelines.
 
==Enclosure==
L-1 1-238Page 42 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.9 Page A-11 First paragraph -new sentenceIn response to RAI 3.3.2.2.5-2, LRA Section A.1.9, "Collection, Drainage, andTreatment Components Inspection Program," is revised to read:A.1.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTSINSPECTION PROGRAMThe Collection, Drainage, and Treatment Components Inspection Programconsists of visual and volumetric inspections. This program will be implementedvia periodic inspections of a representative sample. These inspections willensure that the existing environmental conditions in collection, drainage, andtreatment service are not causing material degradation that could result in a lossof component intended function during the period of extended operation. Visualinspections will be conducted using visual (VT-1 or equivalent) inspectionmethods, capable of detecting loss of material, cracking, or reduction in heattransfer. This program will also include volumetric inspections of inaccessiblesurfaces (e.g., tank bottoms sitting on concrete). The aging effects forelastomers, exposed to raw water, will be monitored through a combination ofvisual inspection and manual or physical manipulation (at least 10 percent ofavailable surface) of the material. Inspections will be performed by qualifiedpersonnel following procedures consistent with the pertinent ASME code ofrecord and 10 CFR 50, Appendix B.
 
==Enclosure==
L-1 1-238Page 43 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.15 Page A-17 Entire sectionIn response to RAI 3.3.2.2.5-2, LRA Section A.1.15, "External SurfacesMonitoring Program," is revised to include physical manipulation of elastomers.Additionally, NRC Region III License Renewal 71002 Inspection Open ItemOIN-352 changes are included in this revision to enhance the program to includeinspection parameters and record retention requirements. LRA Section A.1.15 isreplaced in its entirety to read as follows:A.1.15 EXTERNAL SURFACES MONITORING PROGRAMThe External Surfaces Monitoring Program manages the aging of externalsurfaces, and internal surfaces in cases where environment is the same, ofmechanical components within the scope of license renewal.The External Surfaces Monitoring Program is a condition monitoring pro-gram thatconsists of periodic visual inspections and surveillance activities of componentexternal surfaces to manage cracking and loss of material. The program includescomponents located in Plant systems within the scope of license renewal that areconstructed of aluminum, copper alloy (copper, brass, bronze, and copper-nickel), stainless steel (includinq cast austenitic stainless steel (CASS)). andsteel (carbon and low-alloy steel and cast iron) materials. Crackinq and loss ofmaterial from the external surfaces of these metals will be evidenced by surfaceirregularities, leakage, or localized discoloration and be detectable prior to loss ofintended function. Surfaces that are inaccessible or not readily visible duringeither normal plant operations or refueling outages, such as surfaces that areinsulated, are inspected opportunistically during the period of extendedoperation. Surfaces that are accessible are inspected at a frequency not toexceed one refueling cycle. System inspection and walkdown documentationincludes inspection parameters and acceptance criteria for polymers, elastomersand metallic components as applicable. This documentation is retained in plantrecords.The External Surfaces Monitoring Program, supplemented by the Inspection ofInternal Surfaces in Miscellaneous Piping and Ductinq Program, performsinspection and surveillance of elastomers and Polymers that are exposed to air-indoor uncontrolled and air-outdoor environments, but are not replaced on a setfrequency or interval (i.e., are long-lived), for evidence of cracking, change inmaterial properties (hardening and loss of strength), and loss of material due towear. The aginq effects for elastomers are monitored through a combination of
 
==Enclosure==
L-1 1-238Page 44 of 98visual inspection and manual or physical manipulation (at least 10 percent ofavailable surface) of the material. Acceptance criteria for these componentsconsists of no unacceptable visual indications of cracks or discoloration thatwould lead to loss of function prior to the next scheduled inspection and of nohardening as evidenced by a loss of suppleness during manipulation.The External Surfaces Monitoring Program performs inspection and surveillanceof the CREVS air-cooled condensing unit cooling coil tubes and fins and theSBODG radiator tubes and fins for visible evidence of external surface conditionsthat could result in a reduction in heat transfer. Acceptance criteria for thesecomponents consists of no unacceptable visual indications of foulinq (build up ofdirt or other foreiqn material) that would lead to loss of function prior to the nextscheduled inspection.The External Surfaces Monitorinq Program manages crackinq of copper alloyswith -greater than 15 percent zinc and stainless steel components exposed to anoutdoor air environment through Plant system inspections and walkdowns forevidence of leakage. Acceptance criteria for surfaces consists of nounacceptable visual indications of cracks that would lead to loss of function priorto the next scheduled inspection.
 
==Enclosure==
L-1 1-238Page 45 of 98Affected LRA Section LRA Page No. Affected ParaaraDh and SentenceA. 1.20 Page A-16 Second paragraph, new secondsentenceDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of May 9, 2011, NRC Inspectors requested thatFENOC revise the Fuel Oil Chemistry Program description to state that theprogram monitors and trends water and particulate contamination concentrationsin accordance with the plant's technical specifications, and to include anenhancement to the program to monitor and trend biological activity quarterly.LRA Section A.1.20, "Fuel Oil Chemistry Program," second paragraph, is revisedto read as follows:A.1.20 FUEL OIL CHEMISTRY PROGRAMThe Fuel Oil Chemistry Program manages the presence of contaminants, suchas water or microbiological organisms, that could lead to the onset andpropagation of loss of material or cracking (of susceptible material) throughproper monitoring and control of fuel oil contamination consistent with plantTechnical Specifications and ASTM standards D975, D2276, D2709, D4057 andD4176. Water and particulate contamination concentrations are monitored andtrended in accordance with the plant's Technical Specifications. Biological activityis monitored and trended at least quarterly. The Fuel Oil Chemistry Program is amitigation program.
 
==Enclosure==
L-1 1-238Page 46 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.22 Page A-17 First paragraphIn response to RAI B.2.22-7, the first paragraph of LRA Section A.1.22,"Inservice Inspection (ISI) Program -IWE," is revised to read as follows:A.1.22 INSERVICE INSPECTION (ISI) PROGRAM -IWEThe Inservice Inspection (ISI) Program -IWE establishes responsibilities andrequirements for conducting ASME Code, Section Xl, Subsection IWE (IWE)inspections as required by 10 CFR 50.55a. The Inservice Inspection (ISI)Program -IWE includes examination and testing of accessible surface areas ofthe steel containment; containment hatches and airlocks; seals, gaskets andmoisture barriers; and containment pressure-retaining bolting in accordance withthe requirements of IWE. The program will include examinations to monitor forcracking of containment stainless steel penetration sleeves, dissimilar metalwelds, bellows, and steel components that are subiect to cyclic loadinq but haveno current licensinq basis fatique analysis.
 
==Enclosure==
L-11-238Page 47 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.25 Page A-18 Entire sectionIn response to RAI B.2.25-8 and B.2.39-10, LRA Section A.1.25, "Leak ChaseMonitoring Program," is revised to read as follows:A.1.25 LEAK CHASE MONITORING PROGRAMThe Leak Chase Monitoring Program is a condition monitoring program,consisting of observation and activities to detect leakage from the spent fuel pool,the fuel transfer pit, and the cask pit liners due to age-related degradation.The Leak Chase Monitoring Program includes periodic monitoring of the spentfuel pool, the fuel transfer pit, and the cask pit liners leak chase system. Periodicmonitoring of leakage from the leak chase system permits early determinationand localization of leakage. In conjunction with the PWR Water ChemistryProgram, and, for the spent fuel pool, Technical Specifications requirements formonitoring spent fuel Pool level, the Leak Chase Monitoring Program is creditedfor managing the loss of material aging effect in the treated borated waterenvironment for the stainless steel spent fuel pool, the fuel transfer pit, and thecask pit liners. Loss of material due to crevice or pitting corrosion can occur atweld seams. The program detects and monitors leakage prior to loss of intendedfunction. Measurement of leakage from any monitoring line exceeding 15milliliters per minute will be documented in a condition report for evaluation andpotential corrective actions. Evaluation will include consideration of morefrequent monitoring.The Leak Chase Monitoring Program includes analysis of the leakage from theleak chase system for PH monthly and for iron every six months. The initialacceptance criteria is 7. 0 to 8. 0 for pH. The results for iron are monitored andtrended to insure that there is no indication of corrosion of the reinforcing bars inthe walls or floor of the pool and pits. An acceptance criterion for the ironanalyses will be developed after three years of measurements. Analyses thatexceed the limits will be documented in the Corrective Action Program.The leak chase system preventive maintenance (PM) activity to inspect andclean the leakage pathways is Performed every 18 months based on plant-specific operating experience. Additionally, the program requires inspectionsonce per year of the accessible outside walls and floor (from the ceiling side) ofthe Pool and pits. This inspection will be a documented inspection performed withthe specific intent of identifying indications of leakage migrating through thewalls. Indication of leakage through the walls will be documented in theCorrective Action Program.
 
==Enclosure==
L-1 1-238Page 48 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.28 Page A-19 Second paragraph, last sentenceIn response to RAI B.2.28-1, the second paragraph of LRA Section A.1.28,"Nickel-Alloy Management Program," is revised to read as follows:A.1.28 NICKEL-ALLOY MANAGEMENT PROGRAMThe Nickel-Alloy Management Program uses a number of inspection techniquesto detect cracking, including volumetric and bare metal visual examinations. TheNickel-Alloy Management Program implements the inspections of componentsthrough the Inservice Inspection Program. Component evaluations, examinationmethods, scheduling, and site documentation comply with 10 CFR 50, the ASMECode, NRC bulletins and generic letters, and staff-approved industry guidelinesrelated to nickel-alloy issues. Inspection of dissimilar metal butt welds areconducted in accordance with the requirements of ASME Code Case N-770-1,'Alternative Examination Requirements and Acceptance Standards for Class IPWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 orUNS W86182 Weld Filler Material With or Without Application of Listed MitigationActivities, Section X1, Division 1," as modified by the Code of FederalRegulations, 10 CFR 50.55a Section (q)(6)(ii)(F).
 
==Enclosure==
L-1 1-238Page 49 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.41 Page A-50 Entire sectionIn response to RAIs B.2.41-1 and B.2.41-3, LRA Section A.1.41, "Inspection ofInternal Surfaces in Miscellaneous Piping and Ducting Program," is replaced inits entirety to read as follows:A.1.41 INSPECTION OF INTERNAL SURFACES IN MISCELLANEOUS PIPINGAND DUCTING PROGRAMThe Inspection of Internal Surfaces in Miscellaneous Piping and Ductinq Pro-gramconsists of inspections of the internal surfaces of aluminum, copper alloy(including copper alloy with greater than 15 percent Zn), stainless steel, and steel(includinq gray cast iron) components exposed to air, condensation, dieselexhaust, lubricating oil or moist air: and, external cooling coil surfaces.The program manages loss of material and cracking; loss of material due towear, hardening, and loss of strength of non-metallic, flexible (elastomeric)components: and reduction in heat transfer of coolinq coil tubes and fins.When required by the ASME Code, inspections are conducted in accordancewith the applicable code requirements. In the absence of applicable coderequirements, visual inspections are performed of metallic and polymericcomponent surfaces using plant-specific procedures implemented by inspectorsqualified through plant-specific programs. The inspections are augmented toinclude physical manipulation of non-metallic, flexible (elastomeric) componentsto detect hardening or loss of strength. The sample population for physicalmanipulation is 10 percent of available surface area, including knownsuspect locations.The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Programincludes opportunistic inspections, when components are opened formaintenance, repair, or surveillance to ensure that the existing environmentalconditions are not causing material degradation that could result in a loss ofcomponent intended function during the period of extended operation.Implementation of this program ensures that the intended functions ofsusceptible components are maintained durinq the period of extended operation.
 
==Enclosure==
L-1 1-238Page 50 of 98Parameters Monitored or InspectedAnd Aging Effect for Specific ComponentAcing A Parameter InspectionEffect Mechanism Monitored MethodLoss of Material Crevice Surface Condition, Visual (VT-1 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material Galvanic Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material General Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material MIC Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Pittinq Corrosion Surface Condition, Visual (VT- I or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Erosion Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or U7)Reduction of Foulinq Tube Fouling Visual (VT-3 or equivalent) orHeat Transfer Enhanced VT-1 for CASSCracking SCC or Cyclic Surface Condition. Enhanced Visual (EVT-1 or equivalent) orLoading Cracks Surface Examination (magqnetic particle,liquid penetrant, or Volumetric (RT or UT)(I) When required by the ASME Code, inspections are conducted in accordance with the applicablecode requirements. In the absence of applicable code requirements, visual inspections areperformed of metallic and polymeric component surfaces usinq plant-specific proceduresimplemented by inspectors qualified through plant-specific proaqrams.At least one inspection of each material and environment combination isconducted within the ten-year period prior to enterinq the period ofextended operation.
 
==Enclosure==
L-1 1-238Page 51 of 98Affected LRA Section LRA Page No. Affected Paraaraph and SentenceA.2.5.1 Pages A-44 & Entire sectionA-45In response to RAI 4.6-1, LRA Section A.2.5.1, "Containment Vessel," is revisedto read as follows:A.2.5.1 Containment VesselThe containment vessel is a Class B vessel as defined in the ASME Section III,Paragraph N-132, 1968 Edition through Summer Addenda 1969. Thecontainment vessel meets the requirements for Paragraph N-415.1 of ASMESection III, thereby justifying the exclusion of cyclic or fatigue analyses in thedesign of the containment vessel, as in USAR 3815 TheGontainment vessel has been analyzed for 100 pressure eycles (fro -25 psi tv120 ps, and 400 tem.peratue cycles (fronm 30OF to 120 0F). The oGntainmen.vessel has- not seen any pr-essure cycles in the defined range (thr-ough 2009)'The 60-year proiected cycles for plant heatup and cooldown are 128 (shown inTable 4.3-1) and are less than the specified 400 pressure cycles and400 temperature cycles. Therefore, the values of 400 pressure cycles and400 temperature cycles used to exclude fatigue analyses will not be exceeded for60 years of operation.The TLAA associated with exclusion of the containment vessel from fatigueanalyses per ASME Section III, Paragraph N-415.1 remains valid for the periodof extended operation in accordance with 10 CFR 54.21 (c)(1 )(i).
 
==Enclosure==
L-1 1-238Page 52 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.2.7.5 Page A-50 New SectionA.2.8 Page A-52 New ReferenceAppendix A Table of Page A-5 New entryContentsThe NRC initiated a telephone conference call with FENOC on July 15, 2011, todiscuss the FENOC response to RAI 4.1-2 (ML11172A389). The NRC does notagree that the cycle-dependent fatigue flaw growth analysis is not a time-limitedaging analysis. FENOC provides this supplemental response to RAI 4.6-1 to adda new LRA Section A.2.7.5, "ASME Code Case N-481 Evaluation." LRA SectionA.2.8 is revised to include a new reference to support new Section A.2.7.5. TheLRA Appendix A Table of Contents, not presented below, is revised to includenew Section A.2.7.5. New LRA Section A.2.7.5 and the new reference in LRASection A.2.8 read as follows:A. 275 ASME Code Case N-481 EvaluaticonThe reactor coolant pumps (RCPs) are the only ASME Code Class I pumpsinstalled at Davis-Besse. The pump casings are constructed of cast austeniticstainless steel. The applicable ASME Code for the current Third Ten-YearInspection Interval for Davis-Besse is ASME Section X1, 1995 Edition, throughthe 1996 Addenda, as modified by 10 CFR 50.55a or relief granted inaccordance with 10 CFR 50.55a. Examination Category B-L-1 of this Code yearrequires volumetric examination on pump casing welds. ASME Code CaseN-481, "Alternative Examination Requirements for Cast Austenitic PumpCasings," provides an alternative to the volumetric examination requirement. Thiscode case allows the replacement of volumetric examinations of primary looppump casings with fracture mechanics-based integrity evaluation (Item (d) of thecode case) supplemented by specific visual examinations. Davis-Besse hasinvoked the use of Code Case N-481 in place of the volumetric examinationrequirements of Code Cateqory B-L-1. The NRC has accepted Code Case N-481for use in inservice inspection programs.Code Case N-481 requires an evaluation to demonstrate the safety andserviceability of the pump casings. The evaluation for the Davis-Besse RCPsrequired by Code Case N-481 is documented in Structural Integrity Associates(SIA) report SIR-99-040 [Reference A.2-181. This evaluation assumed a quarterthickness flaw, with length six times its depth, and showed that the flaw willremain stable considering the stresses and material properties of the pumpcasing. To determine stability of the postulated flaw, a fracture mechanics
 
==Enclosure==
L-1 1-238Page 53 of 98evaluation was performed that included a fatique crack growth analysis todemonstrate that a small initial assumed flaw (10 percent through-wall),corresponding to the acceptance standards of ASME Code, Section X1,Subarticle IWB-3500, would not grow to quarter thickness during plant life. Thereare two potential time-dependencies in the Code Case N-481 evaluation.1. The fracture toughness of the cast austenitic stainless steel is not timedependent as the analysis used a lower bound fracture toughness of139 ksiin that bounds the saturated fracture touqhness of theDavis-Besse material2. The fatique crack growth analysis is based on desiqn cycles for a40 year plant life and therefore, is a TLAA requiring analysis anddisposition for license renewalThe fatique crack growth analysis assumed an initial flaw size corresponding tothe acceptance standards of ASME Code Section X1 and considered all thesignificant plant transients. This analysis examined the desiqn cycles anddetermined there were 240 cycles that were significant to flaw growth in theRCPs. Then 2000 cycles were conservatively analyzed, and flaw growth (initial10 percent assumed through-wall had grown only to 15 percent through-wall)remained weli below the quarter thickness postulated flaw. The analyzed cyclesof 2000 bound the 60-year proiected cycles shown in LRA Table 4.3-1 andtherefore, the fatique crack growth TLAA associated with the ASME Code CaseN-481 evaluation will remain valid for the period of extended operation inaccordance with 10 CFR 54.21(c)(1)i).A.2.8 REFERENCESA.2-18 Structural Integrity Associates Report SIR-99-040, "ASME Code CaseN-481, Evaluation of Davis-Besse Reactor Coolant Pumps" Rev. 1,September 2000 (ADAMS Accession No. MLOI 1200090)
 
==Enclosure==
L-1 1-238Page 54 of 98Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable A-1Page A-69Commitment 8In response to RAI 3.3.2.2.5-2, LRA Table A-I, "Davis-Besse License Renewal Commitments," license renewalfuture Commitment 8 is revised to capture revised program enhancements. Additionally, NRC Region III LicenseRenewal 71002 Inspection Open Item OIN-352 changes are included in this revision to enhance the program toinclude inspection parameters and record retention requirements. LRA Table A-i, Commitment 8, is revised to readas follows:Table A-1Davis-Besse License Renewal CommitmentsI I Related LRAItem ImplementationNumber Commitment Schedule Source Section No./Comments8 Enhance the External Surfaces Monitoring Program to: Prior to LRA A. 1.15* Add systems which credit the program for license renewal but April 22, 2017 and B.2.15do not have Maintenance Rule intended functions to the scopeof the program.FENOC Responses toPerform opportunistic inspections of surfaces that are Letters NRC RAIsinaccessible or not readily visible during normal plant operations L-11-153, 3.3.2.2.5-1 andor refueling outages, such as surfaces that are insulated. and B.2.2-2 fromSurfaces that are accessible will be inspected at a frequency not L-11-166 NRC Letterto exceed one refueling cycle. and dated* Perform, in conjunction with the Inspection of Internal Surfaces L-11-238 April 20, 2011,in Miscellaneous Piping and Ducting Program, inspection and andsurveillance of elastomers and polymers exposed to air-indoor NRC RAIuncontrolled or air-outdoor environments, but not replaced on a 3.3.2-2 fromset frequency or interval (i.e., are long-lived), for evidence of NRC Letter
 
==Enclosure==
L-1 1-238Page 55 of 98Table A-1Davis-Besse License Renewal Commitmentsm Ii Related LRANumber Commitment Implementation Source Section No./Number ScheduleCommentscracking and change in material properties (hardening and lossof strength) and loss of material due to wear. Specifyacceptance criteria of no unacceptable visual indications ofcracks or discoloration that would lead to loss of function prior tothe next inspection, and of no hardening as evidenced by a lossof suppleness during manipulation." Perform inspection of the control room emergency ventilationsystem air-cooled condensing unit cooling coil tubes and finsand the station blackout diesel generator radiator tubes and finsfor visible evidence of external surface conditions that couldresult in a reduction in heat transfer. Specify acceptance criteriaof no unacceptable visual indications of fouling (build up of dirtor other foreign material) that would lead to loss of function priorto the next scheduled inspection.* Manage cracking of copper alloys with greater than 15 percentzinc and stainless steel components exposed to an outdoor airenvironment through plant system inspections and walkdownsfor evidence of leakage. Specify acceptance criteria of nounacceptable visual indications of cracks that would lead to lossof function prior to the next scheduled inspection.* Include inspection parameters and acceptance criteria forpolymers. elastomers and metallic components as applicable indatedMay 2, 2011,RAI 3.3.2.2.5-2fromNRC LetterdatedJuly 12, 2011,andNRC OIN-352fromNRC Region Ill71002Inspectionsystem inspection and walkdown documentation. Retain systeminspection and walkdown documentation in plant records.
 
==Enclosure==
L-1 1-238Page 56 of 98Affected LRA SectionLRA Paae No.Affected Paragraph and SentenceTable A-1Page A-69Commitment 26The NRC initiated a telephone conference call with FENOC on July 27, 2011, to discuss the FENOC response toRAI 2.1-3 (ML11126A016). The NRC staff stated that the current FENOC response does not meet the NRC'sexpectations and is not consistent with other applications. FENOC provides this supplemental response toRAI 2.1-3 to replace LRA Table A-i, "Davis-Besse License Renewal Commitments," license renewal futureCommitment 26, in its entirety, to read as follows:Table A-1Davis-Besse License Renewal Commitments______________ I I ___________________I IItemNumberCommitment26 Ensure that abandoned equipment is identified, and either isolated Prior toand drained or included within the scope of license renewal and December 31,subiect to aging management review. 2012Supplementalresponse toNRC RAI2.1-3 fromNRC LetterdatedMarch 30,2011
 
==Enclosure==
L-1 1-238Page 57 of 98Affected LRA SectionLRA Page No.Affected Paraqraph and SentenceTable A-1Page A-55Commitment 28During the NRC Region III Inspection Procedure (IP) 71002, "License Renewal Inspection," held the week ofMay 9, 2011, NRC Inspectors requested that FENOC revise the Fuel Oil Chemistry Program description to state thatthe program monitors and trends water and particulate contamination concentrations in accordance with the plant'stechnical specifications, and to include an enhancement to the program to monitor and trend biological activityquarterly. LRA Table A-I, "Davis-Besse License Renewal Commitments," license renewal future Commitment 28, isrevised to read as follows:Table A-IDavis-Besse License Renewal Commitmentsm Ii Related LRAItem Commitment Implementation Source Section No./Number j Schedule CmetComments28 Enhance the Fuel Oil Chemistry Program to: Prior to LRA A.1.20" Require that internal surfaces of emergency diesel generator April 22, 2017 FENOC B.2.20fuel oil storage tanks and day tanks, diesel oil storage tank, Letter Response todiesel fire pump day tank, and station blackout diesel generator L-1 1-134 NRC RAIday tank are periodically drained (at least once every 10 years) and B.2.20-1 andfor cleaning and are visually inspected to detect potential L-11-238 B.2.20-2 fromdegradation. If degradation is identified in a diesel fuel tank by NRC Lettervisual inspections, a volumetric inspection is performed. dated* Require that biological activity be monitored and trended at least April 5, 2011,quarterly. andNRC OIN-368
 
==Enclosure==
L-1 1-238Page 58 of 98Table A-1Davis-Besse License Renewal CommitmentsMmItem TImplementation Related LRAIter Commitment Sheme Source Section No./Number Schedule CmetCommentsfromNRC Region III71002Inspction
 
==Enclosure==
L-1 1-238Page 59 of 98Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable A-1Page A-69Commitment 30License renewal future Commitment 30 regarding Leak Chase Monitoring Program enhancement is revised basedon the responses to RAI B.2.25-7 and B.2.39-10, and LRA Table A-i, "Davis-Besse License RenewalCommitments," Commitment 30, now reads as follows:Table A-1Davis-Besse License Renewal CommitmentsRelated LRAItem Implementation Source Section No./Number Commitment Schedule Sore ecinN.Comments30 Enhance the Leak Chase Monitoring Program to: Prior to FENOC Response to" Include acceptance criteria such that measurement of leakage April 22, 2017 Letter NRC RAIfrom any monitoring line exceeding 15 milliliters per minute will L-11-153 1.2.25-5 frombe documented in the Corrective Action Program for evaluation L-11-238 datedand potential corrective actions. Evaluation will include April 5, 2011consideration of more frequent monitoring, and* Analyze collected leak chase drainage for PH monthly and for RAts B.2.25iron every six months. The initial acceptance criteria will be 7.0 and B.2.39-10to 8.0 for pH. The results for iron will be monitored and trended fromto insure that there is no indication of corrosion of the reinforcing NRC Letterbars in the walls or floor of the pool and pits. An acceptance datedcriterion for the iron analyses will be developed after three years July 21, 2011of measurements. Analyses that exceed the limits will bedocumented in the Corrective Action Program.* Perform the leak chase inspection and cleaning recurringpreventive maintenance (PM) activity every 18 months.
 
==Enclosure==
L-1 1-238Page 60 of 98Table A-1Davis-Besse License Renewal CommitmentsRelated LRAItem Commitment Implementation Source Section No./Number Schedule CmetCommentsI Inspect once Per year for leakage migrating through theaccessible outside walls and floor (from the ceiling side) of thepool and pits. Document the inspection results and retain inplant records. Indication of leakage through the walls will bedocumented in the Corrective Action Program.
 
==Enclosure==
L-1 1-238Page 61 of 98Affected LRA SectionLRA Page No.Affected ParagraDh and SentenceTable A-1Page A-69Commitment 37License renewal future Commitment 37 regarding core bores for spent fuel pool leakage through concrete isrevised based on the response to RAI B.2.39-10, and LRA Table A-I, "Davis-Besse License RenewalCommitments," Commitment 37, now reads as follows:Table A-1Davis-Besse License Renewal CommitmentsIm Related LRAItem Commitment Implementation Source Section No./Number Schedule CmetComments37 Perform and evaluate core bores of the ECCS Pump Room No. 1 Pr4eF-te FENOC Response towall and the Room 109 ceiling..A9. 2,201 Letters NRC RAIL-1 1-153 B.2.39-2 from* The core bores will be deep enough to expose reinforcing bar in Phase I prior to Letterthe wall and ceiling. The core samples from the core bores will December 31. and datedbe examined for siqns of corrosion or chemical effects of boric 2014 L-11-238 April 5, 2011,acid on the concrete or reinforcing bars. The examination will andinclude a petrographic examination. The reinforcinq steel that RAI B.2.39-10will be exposed for a visual inspection will have corrosion Phase 2 prior to fromproducts collected for testing. Degradation identified from the December 31, NRC Lettersamples will be entered into the FENOC Corrective Action 2020 datedProgram. The core bores will be performed in areas where Julyd21 2011leakage has been observed in the past.* The first set of core bores will be performed prior to the end of2014 (Phase 1).
 
==Enclosure==
L-1 1-238Page 62 of 98Table A-1Davis-Besse License Renewal Commitmentsm Ti Related LRAItem Commitment Implementation Source Section No./Number Schedule -CmetComments" The second set of core bores will be performed prior to the endof 2020 (Phase 2).* Further core bores will be conducted, if warranted, based on theevaluation of the results of the inspection and testing of the corebores or if SFP leakage through the wall or ceiling recurs afterthe second set of core bores is performed. ff spent fuel Poolleakage through another wall or ceiling is identified, then corebores will be performed in a manner similar to that stated for theECCS Pump Room No. 1 wall and the Room 109 ceiling.
 
==Enclosure==
L-1 1-238Page 63 of 98Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable A-1Page A-69Commitment 38License renewal future Commitment 38 regarding concrete cracking on the underside of the spent fuel pool isrevised based on the response to RAI B.2.39-1 0, and LRA Table A-1, "Davis-Besse License RenewalCommitments," Commitment 38, now reads as follows:Table A-1Davis-Besse License Renewal CommitmentsItem IImplementation Related LRANumber Commitment ISchedule Source Section No./Comments38 Evaluate the concrete cracking observed on the underside of the Prior to FENOC Response tospent fuel pool for necessary repairs. !f noscy-, based on tho April 22, 2017 Letter NRC RAIevaluatien, Fopair tho r~ack prior to enteri g the poriod of okdendedL-1-5 B2392fo...,,,-.- ...i ., ......-""&#xf7;'" "'"". .. .n ""+'"'' L-11-153 B.2.39-2 fromopGat~IR, and NRC LetterdatedNote: A core bore of the Room 109 ceiling will be performed by the L-1d-238 April 5, 2011,end of 2014 (see license renewal commitment 37). Degradation andidentified from the samples will be entered into the FENOC RAI B.2.39-10Corrective Action Program. The condition of the concrete and the fromreinforcing steel will be evaluated at that time to assist in NRC Letterdetermining what repairs, if any, need to be made to the underside datedof the spent fuel Pool concrete. The criterion for determining the July21, 2011need to repair the cracking will be the continued capability of thestructures to perform their intended functions during the period ofextended operation.
 
==Enclosure==
L-1 1-238Page 64 of 98Affected LRA SectionLRA Page No.Affected Paragraph and SentenceNew Commitment (Item No. 47)Table A-1Page A-69A new license renewal future commitment is added based on the response to RAI B.2.22-7 regarding examinationof Containment penetrations, and LRA Table A-I, "Davis-Besse License Renewal Commitments," is revised to readas follows:Table A-1Davis-Besse License Renewal CommitmentsI I Related LRAItem Implementation Rected No,Number Commitment Schedule Source Section No.!Comments47 Enhance the Inservice Inspection (ISO) Program -IWE to: Prior to LRA A. 1.22* Include examinations to monitor for cracking of stainless steel April 22, 2017 and B.2.22Containment penetration sleeves, dissimilar metal welds,bellows, and steel components that are subiect to cyclic loadingbut have no current licensing basis fatique analysis. FENOC Response toLetter NRC RAIL- 11-238 B.2.22-7 fromNRC LetterdatedJuly 21, 2011
 
==Enclosure==
L-1 1-238Page 65 of 98Affected LRA SectionLRA Page No.Affected Paraaraah and SentenceTable A-1Page A-69New Commitment (Item No. 48)A new license renewal future commitment is added based on the response to RAI B.2.40-2 regarding theIntake Canal embankment, and LRA Table A-1, "Davis-Besse License Renewal Commitments," is revised toread as follows:Table A-1Davis-Besse License Renewal CommitmentsT Related LRAItem Implementation Source Section No./Number Commitment Schedule commentsComments48 Complete an investigation and needed repairs or modification of the Prior to FENOC Response todegraded portion of the safety-related Intake Canal embankment. April 22, 2017 Letter NRC RAIL-11-238 B.2.40-2 fromNRC LetterdatedJuly 21, 2011
 
==Enclosure==
L-1 1-238Page 66 of 98Affected LRA SectionLRA Page No.Affected Paraqravh and SentenceTable A-1Page A-69New Commitment (Item No. 49)A new license renewal future commitment is added based on the response to RAI B.2.28-1 regarding anenhancement to the Nickel-Alloy Management Program, and LRA Table A-i, "Davis-Besse License RenewalCommitments," is revised to read as follows:Table A-1Davis-Besse License Renewal CommitmentsRelated LRAItem ImplementationRead ANumber Commitment Schedule Source Section No./Comments49 Enhance the Nickel-Alloy Management Program to: Prior to FENOC A. 1.28* Provide for inspection of dissimilar metal butt welds in April 22, 2017 Letter B.2.28accordance with the requirements of ASME Code Case L-11-238 RepnetN-770-1, "Alternative Examination Requirements and NRC RAIAcceptance Standards for Class 1 PWR Piping and Vessel B.2.28-1 fromNozzle Butt Welds Fabricated with UNS N06082 or NRC LetterUNS W86182 Weld Filler Material With or Without Application of datedListed Mitigation Activities, Section X1, Division 1." as modified July 27, 2011by the Code of Federal Regulations, 10 CFR 50.55a(g)(6)(ii)(F).
 
==Enclosure==
L-1 1-238Page 67 of 98Affected LRA SectionLRA Paae No.Pages B-18 &B-21Affected Paragraph and SentenceTable B-23 Rows (as listed)In response to RAI B.2.22-7, the "Inservice Inspection (ISI) Program -IWE" rowof Table B-2, "Consistency of Davis-Besse Aging Management Programs withNUREG-1 801," now shows that enhancement is required, and is revised to readas follows:ConsistentConsistent wtNew / with WitE Plant- EnhancementExisting NUREG- 1801 with Specific Required1801 181wtExceptionsInservice Inspection(ISI) Program -IWE Existing Yes --- YesSection B.2.22In response to RAI B.2.25-8, the "Leak Chase Monitoring Program" row of Table B-2now shows that enhancement is required, and is revised to read as follows:Leak Chase MonitoringProgram Existing Yes YesSection B.2.25In response to RAI B.2.28-1, the "Nickel-Alloy Management Program" row ofTable B-2 now shows that enhancement is required, and is revised to read as follows:Nickel-AlloyManagement Program Existing Yes YesSection B.2.28
 
==Enclosure==
L-11-238Page 68 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.8 Page B-44 Program Description subsection,second paragraphIn response to RAI 3.3.1.49-1, the second paragraph of LRA Section B.2.8,"Closed Cooling Water Chemistry Program," the "Program Description"subsection, previously revised in FENOC Letter dated May 24, 2011(MLI11151A090), is revised to read as follows:B.2.8 CLOSED COOLING WATER CHEMISTRY PROGRAMProgram DescriptionAlso, the Closed Cooling Water Chemistry Program includes corrosion ratemeasurement at selected locations in the closed cooling water systems. Inaddition, periodic inspections of opportunity will be conducted when componentsare opened for maintenance, repair, or surveillance, to ensure that the existingenvironmental conditions are not causing material degradation that could result ina loss of component intended function during the period of extended operation. Arepresentative sample of piping and components will be inspected on a 10-yearinterval, with the first inspection taking place prior to entering the period ofextended operation. Systems within the scope of this Program are monitored forthe presence of microbiological activity in accordance with the EPRI Closed-Cycle Cooling Water guidelines.
 
==Enclosure==
L-1 1-238Page 69 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.9 Page B-47 thru Program Description -newB-49 sentence;Parameters Monitored or Inspectedand Detection of Aging Effects -new paragraphsIn response to RAI 3.3.2.2.5-2, the "Program Description," "ParametersMonitored or Inspected," and "Detection of Aging Effects" subsections of LRASection B.2.9, "Collection, Drainage, and Treatment Components InspectionProgram," are revised to include a new sentence and now read as follows:B.2.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTSINSPECTION PROGRAMProgram DescriptionThe Collection, Drainage, and Treatment Components Inspection Program is anew plant-specific program for Davis-Besse that will consist of visual andvolumetric inspections. This program will be implemented via periodic inspectionsof a representative sample. These inspections will ensure that the existingenvironmental conditions in collection, drainage, and treatment service are notcausing material degradation that could result in a loss of component intendedfunction during the period of extended operation. Visual inspections will beconducted using visual (VT-1 or equivalent) inspection methods, capable ofdetecting loss of material, cracking, or reduction in heat transfer. This programwill also include volumetric inspections of inaccessible surfaces (e.g., tankbottoms sitting on concrete). The aging effects for elastomers, exposed to rawwater, will be monitored through a combination of visual inspection and manualor physical manipulation (at least 10 percent of available surface) of the material.Inspections will be performed by qualified personnel following proceduresconsistent with the pertinent ASME code of record and 10 CFR 50, Appendix B.The Collection, Drainage, and Treatment Components Inspection Program is acondition monitoring program.Aging Management Program Elements* Parameters Monitored or InspectedThe aginq effects for elastomers, exposed to raw water, will be monitoredthrough a combination of visual inspection and manual or physicalmanipulation (at least 10 percent of available surface) of the material
 
==Enclosure==
L-1 1-238Page 70 of 98Detection of Aging EffectsThe aginq effects for elastomers, exposed to raw water, will be monitoredthrouqh a combination of visual inspection and manual or physicalmanivulation (at least 10 Dercent of available surface) of the material.
 
==Enclosure==
L-1 1-238Page 71 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.15 Page B-72 & Program Description andB-73 Enhancements subsectionsIn response to RAI 3.3.2.2.5-2, LRA Section B.2.15, "External SurfacesMonitoring Program," is revised to include physical manipulation of elastomers.Additionally, changes requested in NRC Region III License Renewal 71002Inspection Open Item OIN-352 are included in this revision to enhance theprogram to include inspection parameters and record retention requirements.LRA Section B.2.15, subsections "Program Description" and "Enhancements,"are replaced in their entirety to read as follows:B.2.15 EXTERNAL SURFACES MONITORING PROGRAMProgram DescriptionThe External Surfaces Monitoring Program manages the aging of externalsurfaces, and internal surfaces in cases where environment is the same, ofmechanical components within the scope of license renewal.The External Surfaces Monitoring Pro-gram is a condition monitoring program thatconsists of periodic visual inspections and surveillance activities of componentexternal surfaces to manage cracking and loss of material. The program includescomponents located in plant systems within the scope of license renewal that areconstructed of aluminum, copper alloy (copper, brass, bronze, and copper-nickel), stainless steel (includinq CASS), and steel (carbon and low-alloy steeland cast iron) materials. Cracking and loss of material from the external surfacesof these metals will be evidenced by surface irregularities, leaka-ge, or localizeddiscoloration and be detectable prior to loss of intended function. Surfaces thatare inaccessible or not readily visible during either normal plant operations orrefuelinq outages, such as surfaces that are insulated, will be inspectedopportunistically during the period of extended operation. Surfaces that areaccessible will be inspected at a frequency not to exceed one refueling cycle.System inspection and walkdown documentation will include inspectionparameters and acceptance criteria for polymers, elastomers and metalliccomponents as applicable. This documentation will be retained in plant records.The External Surfaces Monitoring Pro-gram, supplemented by the Inspection ofInternal Surfaces in Miscellaneous Pipinq and Ducting Program, will performinspection and surveillance of elastomers and polymers that are exposed to air-indoor uncontrolled and air-outdoor environments, but are not replaced on a setfrequency or interval (i.e., are Ion-g-lived), for evidence of cracking, change inmaterial properties (hardeninq and loss of strength), and loss of material due to
 
==Enclosure==
L-1 1-238Page 72 of 98wear. The aging effects for elastomers will be monitored through a combinationof visual inspection and manual or physical manipulation (at least 10 percent ofavailable surface) of the material. Acceptance criteria for these components willconsist of no unacceptable visual indications of cracks or discoloration that wouldlead to loss of function prior to the next scheduled inspection hardening asevidenced by a loss of suppleness during manipulation.The External Surfaces Monitoring Program will perform inspection andsurveillance of the CREVS air-cooled condensing unit cooling coil tubes and finsand the SBODG radiator tubes and fins for visible evidence of external surfaceconditions that could result in a reduction in heat transfer. Acceptance criteria forthese components will consist of no unacceptable visual indications of foulinq(build up of dirt or other foreiqn material) that would lead to loss of function priorto the next scheduled inspection.The External Surfaces Monitoring Program will also manage cracking of copperalloys with greater than 15 percent zinc and stainless steel components exposedto an outdoor air environment through plant system inspections and walkdownsfor evidence of leakage. Acceptance criteria for surfaces consists of nounacceptable visual indications of cracks that would lead to loss of function priorto the next scheduled inspection.EnhancementsThe following enhancements will be implemented in the identified programelements prior to the period of extended operation.Scope of Pro-gramSystems that credit the External Surfaces Monitorinq Program for licenserenewal but which do not have Maintenance Rule intended functions willbe added to the scope of the program.Detection of Aging EffectsSurfaces that are inaccessible or not readily visible during either normalplant operations or refueling outages, such as surfaces that are insulated,will be inspected opportunistically during the period of extended operation.Surfaces that are accessible will be inspected at a frequency not toexceed one refueling cycle.
 
==Enclosure==
L-1 1-238Page 73 of 98* Scope of Program, Parameters Monitored/Inspected, Detection ofAging Effects, Acceptance CriteriaThe External Surfaces Monitoring Program, supplemented by theInspection of Internal Surfaces in Miscellaneous Piping and DuctingProgram, will perform inspection and surveillance of elastomers andPolymers exposed to air-indoor uncontrolled or air-outdoor environments,but not replaced on a set frequency or interval (i.e., are long-lived), forevidence of cracking, change in material properties (hardening and loss ofstrength), and loss of material due to wear. The aging effects forelastomers will be monitored through a combination of visual inspectionand manual or physical manipulation (at least 10 percent of availablesurface) of the material. Acceptance criteria for these components willconsist of no unacceptable visual indications of cracks or discoloration thatwould lead to loss of function prior to the next scheduled inspection and ofno hardening as evidenced by a loss of suppleness during manipulation.The External Surfaces Monitoring Program will perform inspection andsurveillance of the CREVS air-cooled condensing unit cooling coil tubesand fins and the SBODG radiator tubes and fins for visible evidence ofexternal surface conditions that could result in a reduction in heat transfer.Acceptance criteria for these components will consist of no unacceptablevisual indications of fouling (build up of dirt or other foreiqn material) thatwould lead to loss of function prior to the next scheduled inspection.The External Surfaces Monitoring Program will also manage cracking ofcopper alloys with greater than 15 percent zinc and stainless steelcomponents exposed to an outdoor air environment through Plant systeminspections and walkdowns for evidence of leakage. Acceptance criteriafor surfaces consists of no unacceptable visual indications of cracks thatwould lead to loss of function prior to the next scheduled inspection.Parameters Monitored/Inspected, Acceptance CriteriaSystem inspection and walkdown documentation will include inspectionparameters and acceptance criteria for polymers, elastomers and metalliccomponents as applicable. This documentation will be retained in plantrecords.
 
==Enclosure==
L-1 1-238Page 74 of 98Affected LRA Section LRA Paae No. Affected Paragraph and SentenceB.2.20 Pages B-87 Program Description -firstand B-88 paragraph, andEnhancements -Monitoring andTrendingDuring the NRC Region III Inspection Procedure (IP) 71002, "License RenewalInspection," held the week of May 9, 2011, NRC Inspectors requested thatFENOC revise the Fuel Oil Chemistry Program description to state that theprogram monitors and trends water and particulate contamination concentrationsin accordance with the plant's technical specifications, and to include anenhancement to the program to monitor and trend biological activity quarterly.LRA Section B.2.20, "Fuel Oil Chemistry Program," subsections "ProgramDescription" and "Enhancements," are revised to read as follows:B.2.20 FUEL OIL CHEMISTRY PROGRAMProgram DescriptionThe Fuel Oil Chemistry Program monitors and maintains fuel oil quality in orderto mitigate damage due to loss of material, as well as due to cracking ofsusceptible materials, for the storage tanks and associated piping andcomponents containing fuel oil that are within the scope of license renewal. Theprogram includes verifying the quality of new fuel oil, periodic sampling of storeddiesel fuel oil, and periodic cleaning and inspection of the emergency dieselgenerator fuel oil storage tanks and day tanks, diesel oil storage tank, diesel firepump day tank, and station blackout diesel generator day tank. The Fuel OilChemistry Program manages the presence of contaminants, such as water ormicrobiological organisms, that could lead to the onset and propagation of loss ofmaterial or cracking (of susceptible material) through proper monitoring andcontrol of fuel oil contamination consistent with plant Technical Specificationsand ASTM standards D975, D2276, D2709, D4057 and D4176. Water andparticulate contamination concentrations are monitored and trended inaccordance with the plant's Technical Specifications. Biological activity will bemonitored and trended at least quarterly. Exposure to these contaminants areminimized by a) verifying the quality of new fuel oil before it enters the storagetanks, b) periodic sampling of tank contents to ensure the fuel oil is free of waterand particulates, and c) periodic cleaning and inspection of tanks containing fueloil. Fuel oil tanks will be periodically drained (at least once every 10 years) forcleaning and will be visually inspected to detect potential degradation. Ifdegradation is identified in a diesel fuel tank by visual inspections, a volumetric
 
==Enclosure==
L-1 1-238Page 75 of 98inspection will be performed. The Fuel Oil Chemistry Program is a mitigationprogram.EnhancementsThe following enhancements will be implemented in the identified programelements prior to the period of extended operation.* Monitoring and TrendingRequire that biological activity be monitored and trended at least quarterly.
 
==Enclosure==
L-1 1-238Page 76 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.22 Page B-96 Program Description subsection,first paragraph; and,Enhancements subsectionIn response to RAI B.2.22-7, the first paragraph of LRA Section B.2.22,"Inservice Inspection (ISI) Program -IWE," "Program Description," is revised, anda new enhancement is added to the program, as follows:B.2.22 INSERVICE INSPECTION (ISI) PROGRAM -IWEProgram DescriptionThe Inservice Inspection (ISI) Program -IWE establishes responsibilities andrequirements for conducting ASME Code Section Xl, Subsection IWE inspectionsas required by 10 CFR 50.55a. The Inservice Inspection (ISI) Program -IWEincludes examination and/or testing of accessible surface areas of the steelcontainment vessel; containment hatches and airlocks; seals, gaskets andmoisture barriers; and containment pressure-retaining bolting. Theseexaminations are in accordance with the requirements of the ASME Code,Section Xl, 1995 Edition through the 1996 Addenda. The program will includeexaminations to monitor for cracking of Containment stainless steel penetrationsleeves, dissimilar metal welds, bellows, and steel components that are subiectto cyclic loadinq but have no current licensing basis fatique analysis.EnhancementsThe following enhancement will be implemented in the identified programelement prior to the period of extended operation.* Parameters Monitored or InspectedThe Inservice Inspection (ISI) Program -IWE will include examinations tomonitor for cracking of Containment stainless steel penetration sleeves,dissimilar metal welds, bellows, and steel components that are subiect tocyclic loading but have no current licensing basis fatique analysis.
 
==Enclosure==
L-11-238Page 77 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.25 Pages B-104 Program Description;thru B-107 Parameters Monitored or Inspected;Detection of Aging Effects;Acceptance Criteria; and,EnhancementsIn response to RAI B.2.25-8 and B.2.39-10, the "Program Description,""Acceptance Criteria," and "Enhancements" subsections of LRA Section B.2.25,"Leak Chase Monitoring Program," subsection, are revised as follows:B.2.25 LEAK CHASE MONITORING PROGRAMProgram DescriptionThe Leak Chase Monitoring Program is an existing condition monitoring program,consisting of observation and activities to detect leakage from the spent fuel pool,the fuel transfer pit, and the cask pit liners due to age-related degradation.The Leak Chase Monitoring Program includes periodic monitoring of the spentfuel pool, the fuel transfer pit, and the cask pit liners leak chase system. Periodicmonitoring of leakage from the leak chase system permits early determinationand localization of leakage. In conjunction with the PWR Water ChemistryProgram, and, for the spent fuel pool, Technical Specifications requirements formonitoring spent fuel pool level, the Leak Chase Monitoring Program is creditedfor mana-ging the loss of material aging effect in the treated borated waterenvironment for the stainless steel spent fuel pool, the fuel transfer pit, and thecask pit liners. Loss of material due to crevice or pitting corrosion can occur atweld seams. The program detects and monitors leakage prior to loss of intendedfunction. Measurement of Ieaka-e from any monitorinq ine exceeding 15milliliters per minute will be documented in a condition report for evaluation andpotential corrective actions. Evaluation will include consideration of morefrequent monitoring.The Leak Chase Monitoring Program will include analysis of the leakage from theleak chase system for pH monthly and for iron every six months. The initialacceptance criteria will be 7. 0 to 8. 0 for pH. The results for iron will be monitoredand trended to insure that there is no indication of corrosion of the reinforcingbars in the walls or floor of the pool and pits. An acceptance criterion for the ironanalyses will be developed after three years of measurements. Analyses thatexceed the limits will be documented in the Corrective Action Program.The leak chase system recurring preventive maintenance (PM) activity to inspectand clean the leakage nathwavs will be performed every 18 months based on
 
==Enclosure==
L-11-238Page 78 of 98plant-specific operating experience. Additionally, the program will requireinspections once per year of the accessible outside walls and floor (from theceiling side) of the pool and pits. This inspection will be a documented inspectionperformed with the specific intent of identifying indications of leakage migratinqthrough the walls. Indication of leakage through the walls will be documented inthe Corrective Action Program.Aging Management Program ElementsThe results of an evaluation of each program element are provided below.* Parameters Monitored or InspectedThe spent fuel pool, the fuel transfer pit, and the cask pit liner leak detectiondrain valves are periodically opened, any leakage is collected, and theamounts are recorded. In addition, leak rates for zone valves are calculatedby the volumetric method and recorded. The Leak Chase Monitoring Programincludes analysis of the leakage from the leak chase system for DH monthlyand for iron every six months. The results for iron will be monitored andtrended to insure that there is no indication of corrosion of the reinforcing barsin the walls or floor of the pool and pits. Additionally, the program requiresinspections once per year of the accessible outside walls and floor (from theceiling side) of the pool and pits, with the specific intent of identifyingindications of leakage migratinq through the walls." Detection of Aging EffectsThe Leak Chase Monitoring Program includes activities to cycle open andclose the spent fuel pool, the fuel transfer pit, and the cask pit liner drainvalves on a monthly basis [References LC.3a Section 4.1 and LC.4a]. Eachvalve on the drain line capable of being cycled is opened to allow any waterthat accumulated in the lines to drain into an open funnel. After a prescribedwait time, leakage is collected. The amount collected and the calculated leakrate are recorded for each of the 21 drain zones. If leakage collected fromany zone drain valve is greater than 10 ml, then the sample is appropriatelylabeled and transported to a laboratory for boron analysis. Collected leakageinformation and boron analysis results are recorded in the work order system.The Leak Chase Monitoring Program includes analysis of the leakage fromthe leak chase system for oH monthly and for iron every six months.Monitoring of leakage from the leak chase system permits early determinationand localization of any leakage.The leak chase system preventive maintenance (PM) activity to inspect andclean the leakage pathways is performed every 18 months based on plant-
 
==Enclosure==
L-11-238Page 79 of 98specific operating experience. Additionally, the program requires inspectionsonce per year of the accessible outside walls and floor (from the ceiling side)of the pool and pits. This inspection will be performed with the specific intentof identifying indications of leakage m4gratinq through the walls. Theinspection results will be documented and retained in plant records.Acceptance CriteriaMeasurement of leakage from any monitorinq line exceedinq 15 ml/min will bedocumented in the Corrective Action Program for evaluation and potentialcorrective actions. Evaluation will include consideration of more frequentmonitoring. Adverse trends (continued increases of leak rates on a particularzone valve) are also documented in the Corrective Action Program.EnhancementsThe following enhancements will be implemented in the identified programelements prior to the period of extended operation.* Parameters Monitored or Inspected, Detection of Agina Effects andAcceptance CriteriaAnalyze collected leak chase drainage for PH monthly and for iron everysix months. The initial acceptance criteria will be 7.0 to 8.0 for pH. Theresults for iron will be monitored and trended to insure that there is noindication of corrosion of the reinforcing bars in the walls or floor of thePool and pits. An acceptance criterion for the iron analyses will bedeveloped after three years of measurements. Analyses that exceed thelimits will be documented in the Corrective Action Program.Inspect once per year for leakage migrating through the accessibleoutside walls and floor (from the ceilinq side) of the Pool and pits. Theacceptance criterion is no visible leakage. Document the inspection resultsand retain in plant records. Indication of leakage throuqh the walls will bedocumented in the Corrective Action Program." Detection of Aging EffectsPerform the leak chase inspection and cleaning recurrinq preventivemaintenance (PM) activity every 18 months.
 
==Enclosure==
L-1 1-238Page 80 of 98Acceptance CriteriaInclude acceptance criteria such that measurement of leakage from anymonitoring line exceeding 15 ml/min will be documented in the CorrectiveAction Program for evaluation and potential corrective actions. Evaluationwill include consideration of more frequent monitoring.
 
==Enclosure==
L-1 1-238Page 81 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.28 Pages B-113 Program Description -new thirdthru B-115 paragraph;Detection of Aging Effects -newfourth paragraph;Monitoring and Trending -newthird paragraph; and,Enhancements -new enhancementIn response to RAI B.2.28-1, the "Program Description," "Detection of AgingEffects," "Monitoring and Trending," and "Enhancements" subsections ofLRA Section B.2.28, "Nickel-Alloy Management Program," are revised as follows:B.2.28 NICKEL-ALLOY MANAGEMENT PROGRAMProgram DescriptionIn addition, inspection of dissimilar metal butt welds will be conducted inaccordance with the requirements of ASME Code Case N-770-1, 'AlternativeExamination Requirements and Acceptance Standards for Class I PWR Pipingand Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182Weld Filler Material With or Without Application of Listed Mitigation Activities,Section X1, Division 1," as modified by the Code of Federal Regulations,10 CFR 50.55a Section (g)(6)(ii)(F).Aging Management Program ElementsThe results of an evaluation of each program element are provided below.Detection of Aging EffectsIn addition, inspection of dissimilar metal butt welds will be conducted inaccordance with the requirements of ASME Code Case N-770-1, 'AlternativeExamination Requirements and Acceptance Standards for Class 1 PWRPiping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 orUNS W86182 Weld Filler Material With or Without Application of ListedMitigation Activities, Section X1, Division 1," as modified by the Code ofFederal Regulations, 10 CFR 50.55a Section (g)(6)(ii)(F).
 
==Enclosure==
L-1 1-238Page 82 of 98Monitoring and TrendingIn addition, MRP-139 will be replaced by incorporating the requirements ofASME Code Case N-770 as modified by the Code of Federal Regulations,10 CFR 50.55a Section (c)(6)(ii)(F).EnhancementsThe following enhancement will be implemented in the identified programelements prior to the period of extended operation.* Detection of Aging Effects. Monitoring and TrendingProvide for inspection of dissimilar metal butt welds in accordance with therequirements of ASME Code Case N-770-1, "Alternative ExaminationRequirements and Acceptance Standards for Class I PWR Piping andVessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182Weld Filler Material With or Without Application of Listed MitigationActivities, Section X1, Division 1," as modified by the Code of FederalRegulations, 10 CFR 50.55a Section (g)(6)(ii)(F).
 
==Enclosure==
L-1 1-238Page 83 of 98Affected LRA SectionLRA Page No.Page B-123Affected ParaoraDh and SentenceB.2.30Aging Management Program Element:"Detection of Aging Effects"The NRC initiated a telephone conference call with FENOC on July 27, 2011, todiscuss FENOC's response to RAI 3.3.2.2.4.3-1 and corresponding LRASection B.2.30 amendment (MLI 11159A1 32). The NRC requests that FENOCidentify the type of visual examination that will be used to inspect for cracking aspart of the One-Time Inspection Program. FENOC provides this supplementalresponse to RAI 3.3.2.2.4.3-1 to revise LRA Section B.2.30, "One-TimeInspection," Aging Management Program Element "Detection of Aging Effects,"to include a table identifying the types of inspections that are planned to beperformed, as follows:B.2.30 ONE-TIME INSPECTIONAging Management Program Elements0 Detection of Aging EffectsParameters Monitored or InspectedAnd A-ging Effect for Specific ComponentAging Aging Parameter InspectionEffect Mechanism Monitored Method "'Loss of Material Crevice Surface Condition, Visual (VT-1 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material Galvanic Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material General Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material MIC Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Pitting Corrosion Surface Condition, Visual (VT- I or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Erosion Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)
 
==Enclosure==
L-1 1-238Page 84 of 98Parameters Monitored or InspectedAnd A-ging Effect for Specific Component, cont.Agin Aging Parameter InspectionEffect Mechanism Monitored Method "IReduction of Fouling Tube Fouling Visual (VT-3 or equivalent) orHeat Transfer Enhanced VT-1 for CASSCracking SCC or Cyclic Surface Condition, Enhanced Visual (EVT-1 or equivalent) orLoading Cracks Surface Examination (magnetic particle,liquid penetrant, or Volumetric (RT or UT)(I) Examinations of code components will follow procedures consistent with the requirements of theASME Code and 10 CFR 50 Appendix B. Non-code examinations will be performed in accordancewith site procedures.
 
==Enclosure==
L-1 1-238Page 85 of 98Affected LRA Section LRA Paace No. Affected ParaclraDh and SentenceB.2.41 Page B-166 Program Description -entire section;Scope -third paragraph;Parameters Monitored or Inspected -new table; and,Detection of Aging Effects -secondparagraphIn response to RAIs B.2.41-1 and B.2.41-3, LRA Section B.2.41, "Inspection ofInternal Surfaces in Miscellaneous Piping and Ducting Program," subsections"Program Description," "Scope," "Parameters Monitored or Inspected," and"Detection of Aging Effects," are revised as follows:B.2.41 INSPECTION OF INTERNAL SURFACES IN MISCELLANEOUS PIPINGAND DUCTING PROGRAMProgram DescriptionThe Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Programis a new plant-specific program for Davis-Besse. The program will consist ofinspections of the internal surfaces of aluminum, copper alloy (including copperalloy > 15% Zn), stainless steel, and steel (including gray cast iron) componentsexposed to air, condensation, diesel exhaust, lubricating oil or moist air; andexternal cooling coil surfaces. The hnection of Internal .. inMiscllaneus Piping and Ducting Pr, gram wi also consist of accessibleinsections of tMe intornal surfaces of nogn meta~h'c, flexible (elastOmger4G)components that ara not included in other- 0gn mangement progr-ams and theexternal of non met-llic, , lastornerc components as -asupplement to the External Sur-acos Monitorng Proegram.The program will manage loss of material and cracking, of ...ceptible stainlesssteel components, loss of material due to wear, hardening and loss of strength ofnon-metallic, flexible (elastomeric) components; and reduction in heat transfer ofcooling coil tubes and fins.When required by the ASME Code, inspections are conducted in accordancewith the applicable code requirements. In the absence of applicable coderequirements, visual inspections are performed of metallic and polymericcomponent surfaces usinq plant-specific procedures implemented by inspectorsqualified throuqh plant-specific pro-qrams. The inspections will be auqmented toinclude physical manipulation of non-metallic, flexible (elastomeric) components
 
==Enclosure==
L-11-238Page 86 of 98to detect hardeninq or loss of strength. The sample population for physicalmanipulation will be 10 percent of available surface area. including knownsuspect locations.The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Programwill include opportunistic inspections, when components are opened formaintenance, repair, or surveillance to ensure that the existing environmentalconditions are not causing material degradation that could result in a loss ofcomponent intended function during the period of extended operation.Implementation of this program will ensure that the intended functions ofsusceptible components are maintained during the period of extended operation.Parameters Monitored or InspectedAnd Aaina Effect for SDecific ComDonentAging A"ing Parameter InspectionEffect Mechanism Monitored Method "ILoss of Material Crevice Surface Condition, Visual (VT- I or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT2Loss of Material Galvanic Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material General Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material MIC Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Pitting Corrosion Surface Condition, Visual (VT- I or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Erosion Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Reduction of Foulin_ Tube Foulinq Visual (VT-3 or equivalent) orHeat Transfer Enhanced VT-1 for CASSCracking SCC or Cyclic Surface Condition, Enhanced Visual (EVT-I or equivalent) orLoading Cracks Surface Examination (magnetic particle,liquid penetrant, or Volumetric (RT or UT)(1) When required by the ASME Code, inspections are conducted in accordance with the applicablecode requirements. In the absence of applicable code requirements, visual inspections areperformed of metallic and polymeric component surfaces usinq plant-specific proceduresimplemented by inspectors qualified through plant-specific programs.The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Programis a new condition-monitoring program. At least one inspection of each materialand environment combination will be conducted within the 10-year period prior toentering the period of extended operation.
 
==Enclosure==
L-1 1-238Page 87 of 98Aging Management Program ElementsThe results of an evaluation of each program element are provided below.ScopeThe program will include visual and physical (manipulation or prodding)examination of subject non-metallic, flexible (elastomeric) components invarious environments for evidence of hardening or loss of strength due tothermal exposure, ultraviolet exposure, or ionizing radiation as a s-pp!ement#^~ f*h% Qvt~rn Q, ,fn, ,&#xfd; A Arnr i~fr~nryr, D2~r_.-u",M0 Parameters Monitored or InspectedParameters Monitored or InspectedA nA A nirn =a^ ftu' Qnn-, f_'rMnrftnfnf"lu .. --M.,#M I.~.# I w.#S S .S S** v e lU*UI'AeISS*A"ing Afing Parameter InspectionEffect Mechanism Monitored Method "ILoss of Material Crevice Surface Condition, Visual (VT-1 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material Galvanic Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetrc (RT or UT)Loss of Material General Surface Condition, Visual (VT-3 or equivalent) and/orCorrosion Wall Thickness Volumetric (RT or UT)Loss of Material MIC Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Pitting Corrosion Surface Condition, Visual (VT-I or equivalent) and/orWall Thickness Volumetric (RT or UT)Loss of Material Erosion Surface Condition, Visual (VT-3 or equivalent) and/orWall Thickness Volumetric (RT or UT)Reduction of Foulinq Tube Fouling Visual (VT-3 or equivalent) orHeat Transfer Enhanced VT-1 for CASSCrackinq SCC or Cyclic Surface Condition, Enhanced Visual (EVT-1 or equivalent) orLoadinq Cracks Surface Examination (magnetic particle,liquid penetrant, or Volumetric (RT or UT)(I) When required by the ASME Code, inspections are conducted in accordance with the applicablecode requirements. In the absence of applicable code requirements, visual inspections areperformed of metallic and polymeric component surfaces using plant-specific proceduresimplemented by inspectors qualified through plant-specific programs.
 
==Enclosure==
L-1 1-238Page 88 of 98Detection of Aging EffectsWhen required by the ASME Code, inspections are conducted in accordancewith the applicable code requirements. In the absence of applicable coderequirements, visual inspections are performed of metallic and polymericcomponent surfaces using plant specific procedures implemented byinspectors qualified through plant-specific pro-grams.
 
==Enclosure==
L-1 1-238Page 89 of 98Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.42 Page B-166 Entire SectionThe NRC initiated a telephone conference call with FENOC on July 27, 2011, todiscuss the FENOC response to RAI XI.S8-1 (ML1 1172A389). The NRC staffasked for clarification as to why ASTM D5163 specifies a year-of-issuedesignator in part of the submittal but is not consistent throughout the response.FENOC provides this supplemental response to RAI XI.S8-1 to add year-of-issuedesignators to the ASTM standards cited in the program. LRA Section B.2.42,"Nuclear Safety-Related Coatings Program," is revised accordingly (sixlocations), and now reads as follows:B.2.42 NUCLEAR SAFETY-RELATED COATINGS PROGRAMProgram DescriptionThe Nuclear Safety-Related Protective Coatings Program is an existing plant-specific condition monitoring program that monitors the performance of ServiceLevel 1 coatings inside containment (e.g., coated structures and componentssuch as steel containment vessel, structural steel, supports, penetrations, andconcrete walls and floors) through periodic coating examinations, conditionassessments and remedial actions, including repair or testing. The NuclearSafety-Related Protective Coatings Program defines roles, responsibilities,controls and deliverables for monitoring the condition of coatings in containment.Service Level 1 coatings are subject to the guidance of ASTM International(ASTM) D5163-91, "Standard Guide for Establishing Procedures to Monitor thePerformance of Safety Related Coatings in an Operating Nuclear Power Plant,"and American National Standards Institute (ANSI) Standard N101.4 (1972),"Quality Assurance for Protective Coatings Applied to Nuclear Facilities." Theprogram follows the guidance of EPRI 1003102, "Guidelines on Nuclear SafetyRelated Coatings," Revision 1. This program also ensures that the Design BasisAccident (DBA) analysis limits with regard to debris loading from failed coatingswill not be exceeded for the Emergency Core Cooling Systems (ECCS) suctionstrainers. On July 14, 1998 the NRC published Generic Letter 98-04, "Potentialfor Degradation of the Emergency Core Cooling System and the ContainmentSpray System After a Loss-of-Coolant Accident because of Construction andProtective Coating Deficiencies and Foreign Material in Containment." Theprogram is implemented as described in the FirstEnergy Nuclear OperatingCompany (FENOC) response to NRC Generic Letter 98-04, accepted by theNRC. The Nuclear Safety-Related Protective Coatings Program providesreasonable assurance that potentially detrimental aging effects will be adequatelydetected and mitigated such that Service Level 1 protective coatings are
 
==Enclosure==
L-11-238Page 90 of 98maintained consistent with the current licensing basis for the period of extendedoperation.NUREG-1801 ConsistencyThe Nuclear Safety-Related Protective Coatings Program is an existing plantspecific program for Davis-Besse. While NUREG-1801 includes a ProtectiveCoating Monitoring and Maintenance Program (XI.S8), the Nuclear Safety-Related Protective Coatings Program is considered plant-specific, and isevaluated against the ten elements described in Appendix A.1, Section A.1.2.3 ofNUREG-1800, the Standard Review Plan for License Renewal (SRP-LR).Aging Management Program ElementsThe results of an evaluation of each program element are provided below.* ScopeThe Nuclear Safety-Related Protective Coatings Program monitors theperformance of Service Level 1 coatings inside containment through periodiccoating examinations, condition assessments and remedial actions, includingrepair or testing. The Nuclear Safety-Related Protective Coatings Programensures that the Design Basis Accident (DBA) analysis limits with regard tocoatings will not be exceeded for the ECCS suction strainers per theresponse to NRC Generic Letter 98-04. The program consists of periodicvisual inspections of the Service Level 1 coatings, looking for any visibledefects, such as blistering, cracking, flaking, peeling, delamination, rustingand physical damage. The program was established in accordance with theguidance provided in ASTM D5163-91 "Standard Guide for EstablishingProcedures to Monitor the Performance of Safety Related Coatings in anOperating Nuclear Power Plant."The qualification testing of Service Level 1 coatings used for new applicationsor used as maintenance coatings for repair and replacement activities insidecontainment is addressed in the FENOC revised response to NRC GenericLetter 98-04 for Davis-Besse. The testing meets the applicable requirementscontained in Regulatory Guide (RG) 1.54 Rev. 0, "Quality AssuranceRequirements for Protective Coatings Applied to Water-Cooled NuclearPower Plants." Although Davis-Besse was not committed to ANSI N101.2,"Protective Coatings (Paints) for Light Water Nuclear Reactor ContainmentFacilities," protective coatings have been evaluated to meet the coatingsqualification test criteria per ANSI N101.2.
 
==Enclosure==
L-1 1-238Page 91 of 98" Preventive ActionsProtective coatings are not credited for aging management at Davis-Besse.The Nuclear Safety-Related Protective Coatings Program is a conditionmonitoring program that does not include preventive actions. No actions aretaken as part of the Nuclear Safety-Related Protective Coatings Program toprevent aging effects or mitigate age-related degradation." Parameters Monitored or InspectedThe Nuclear Safety-Related Protective Coatings Program monitors ServiceLevel I coatings in accordance with ASTM D5163-91, "Standard Guide forEstablishing Procedures to Monitor the Performance of Safety RelatedCoatings in an Operating Nuclear Power Plant," ASTM D 714-0 2 "StandardTest method for Evaluating Degree of Blistering of Paints," and SSPC VIS-2,"Standard Method of Evaluating Degree of Rusting on Painted Surfaces."Parameters monitored or inspected by the Nuclear Safety-Related ProtectiveCoatings Program include any visible defects, such as blistering, cracking,flaking, peeling, delamination, rusting and physical damage.The Nuclear Safety-Related Protective Coatings Program procedure will berevised to clarify that visible defects "rusting and physical damage" areinspection attributes following the guidance of ASTM D5163-08,subparagraph 10.2. The Coating Condition Assessment Inspection Form willbe revised to list the same set of degradation parameters for inspection as thegoverning procedure." Detection of Aging EffectsA visual containment inspection is performed for evidence of degradedqualified coatings and identification of unqualified coatings applied tostructures and components during each refueling outage in accordance withthe guidance in ASTM D5163-91, "Standard Guide for EstablishingProcedures to Monitor the Performance of Safety Related Coatings in anOperating Nuclear Power Plant." The containment inspection includes avisual coating inspection of the accessible areas that are listed in theapproved procedure along with location plan maps. Unless conditions warranta closer review, inspectors are not required to examine portions of the area,structures or components that are inaccessible due to insulation, scaffold orpermanent plant SSCs. Conditions that warrant a closer review are evidenceof a coating failure where the area of concern is hidden from view by theobstruction. For areas of the Containment Vessel which have visual evidence(identifiable boundary) of repair or touch-up; its location (azimuth andelevation), approximate surface area and average dry film thickness are
 
==Enclosure==
L-11-238Page 92 of 98documented on the Coating Condition Assessment Inspection Form.Instruments and equipment used for inspection; such as flashlight, acuitycard, inspection mirror, camera, telescope, video equipment, magnifyingglass, measuring tape, dry film thickness gage, spring micrometer, etc. meetthe guidelines of ASTM D5163-08, subparagraph 10.5.Coating inspections are performed by coatings inspectors qualified per therequirements of Regulatory Guide 1.58, "Qualification of Nuclear Power PlantInspection, Examination and Testing Personnel," and ANSI N45.2.6,"Qualification of Inspection, Examination, and Testing Personnel for NuclearPower Plants". The nuclear safety-related coatings program owner andcoating surveillance personnel meet the requirements of EPRI 1003102Revision 1, "Guidelines on Nuclear Safety Related Coatings."The Nuclear Safety-Related Protective Coatings Program procedure will berevised to specify the qualifications for inspection personnel, the inspectioncoordinator and the inspection results evaluator following the guidance ofASTM D5163-08, paragraph 9.Monitoring and TrendingThe Nuclear Safety-Related Protective Coatings Program incorporatesguidance from ASTM D5163-91, "Standard Guide for Establishing Proceduresto Monitor the Performance of Safety Related Coatings in an OperatingNuclear Power Plant." The Nuclear Safety-Related Coatings Program ownerdevelops and manages the Nuclear Safety-Related Protective CoatingsProgram. The Nuclear Safety-Related Coatings Program owner alsomaintains the Non-DBA Qualified Protective Coatings Inventory. Inspectionresults are reviewed and identified degradations are evaluated in accordancewith the FENOC Corrective Action Program. Degraded coating that is left inplace in an area is documented on the Coating Condition AssessmentInspection form and evaluated by the program owner.The Nuclear Safety-Related Protective Coatings Program procedure will berevised to include prioritization of repair areas as either needing repair duringthe same outage or as postponed to future outages, but under surveillance inthe interim period, following the guidance of ASTM D5163-08, subparagraph11.1.2." Acceptance CriteriaThe Nuclear Safety-Related Protective Coatings Program characterizes,documents, and tests defective or deficient coatings in accordance withASTM D5163-91 "Standard Guide for Establishing Procedures to Monitor thePerformance of Safety Related Coatings in an Operating Nuclear Power
 
==Enclosure==
L-1 1-238Page 93 of 98Plant." As applicable, coated surfaces are characterized as exhibiting blisters,cracking, flaking, peeling, delamination, abrasion, and holidays. Coating testsare employed for areas where the qualification is in question, representativedry film thickness is obtained for each area on a structure or componentwhich has coating degradation. Evidence of corrosion is further categorizedper the guidance of a standard method for evaluating degree of rusting onpainted surfaces.The Coating Surveillance Personnel inspect the containment according to thefollowing degradation definitions:* Abrasion -The wearing away of coating material in small shreds as aresult of friction." Blistering -The formation of bubbles in a cured, or nearly cured,coating film after exposure, generally in an aqueous environment." Cracking -The formation of breaks in a coating film that extendthrough to the underlying surface.* Delamination -A separation of one coat from another coat within acoating system; or from the substrate." Flaking -The detachment of small pieces of the coating film." Holiday -Pinhole, skip, discontinuity, or void in a coating film thatexposes the substrate." Peeling -The separation of one or more coats or layers of a coatingsystem from the substrate.Acceptable coatings are free of delamination, blistering, peeling, flaking,cracking and other defects. Coatings not found to be acceptable aredocumented using the FENOC Corrective Action Program. The protectivecoating condition assessment and associated Coating Condition AssessmentInspection forms are approved and signed by the Protective CoatingsProgram Owner or their Designee.The Nuclear Safety-Related Protective Coatings Program procedure will berevised to improve reporting requirements by following the guidance ofASTM D5163-08, paragraph 11, including a summary report of findings andrecommendations for future surveillance or repair, and prioritization of repairs.
 
==Enclosure==
L-1 1-238Page 94 of 98" Corrective ActionsThis element is common to Davis-Besse programs and activities that arecredited with aging management during the period of extended operation andis discussed in Section B.1.3.* Confirmation ProcessThis element is common to Davis-Besse programs and activities that arecredited with aging management during the period of extended operation andis discussed in Section B.1.3.* Administrative ControlsThis element is common to Davis-Besse programns and activities that arecredited with aging management during the period of extended operation andis discussed in Section B.1.3.* Operating ExperienceA review of operating experience indicates that the Nuclear Safety-RelatedProtective Coatings Program has been effective in monitoring coatings insidecontainment by identifying degraded conditions, performing evaluations andperforming corrective actions ensuring that the DBA analysis limits for debrisloading will not be exceeded for the ECCS suction strainers.Industry operating experience is documented in NRC Regulatory Guide 1.54and several NRC Generic Communications including Information Notice97-13, Generic Letter 98-04, Bulletin 2003-01 and Generic Letter 2004-02,"Potential Impact of Debris Blockage on Emergency Recirculation duringDesign Basis Accidents at Pressurized-Water Reactors."The industry experience cited in these publications deals principally withdebris that could block emergency recirculation during a design basisaccident.In 2003 Davis-Besse provided a revised response to NRC GenericLetter 98-04. During the Cycle 13 refueling outage, FENOC identified via theCorrective Action Program that significant amounts of unqualified coatingmaterials were applied to components inside the containment vessel. FENOCinformed the NRC by letter dated September 15, 2003 that incomplete orinaccurate information was provided in the original 1998 Davis-Besseresponse to Generic Letter 98-04. This issue led to reporting that thecontainment emergency sump could be significantly challenged by thequantity of failed coating material and other debris present in the Containment
 
==Enclosure==
L-1 1-238Page 95 of 98after a postulated Loss of Coolant Accident (LOCA) under Davis-BesseLicensee Event Report (LER) 2002-005. Corrective actions taken for thisevent were:" The old Containment Emergency Sump Strainer was removed and anew strainer with greater surface area was installed." Unqualified coatings have been removed from major equipment inContainment and replaced with qualified coatings." A Nuclear Safety-Related Coatings Program has been developed forcoating material controls and application to structures and componentslocated within the Containment." Where possible, fibrous insulation was removed from Containment.The fibrous insulation and unqualified coatings left in the Containmenthave been identified and evaluated (in conjunction with other potentialdebris) for effect on the Emergency Core Cooling System andContainment Spray System. Controls have been established forpotential debris sources to ensure requirements are met.* Evaluations were performed in conjunction with the modificationsimplemented on the containment emergency sump, which examinedthe Low Pressure Injection System, the High Pressure InjectionSystem, the Containment Spray System, and the Boron PrecipitationControl System." Modifications were implemented for the High Pressure InjectionSystem Pumps.In 2004, the NRC concluded that information regarding the reason for theviolation based on the FENOC November 11, 1998 response to GenericLetter 98-04, the corrective actions taken, plans to correct the violation andprevent recurrence, and the date when full compliance was achieved, wereadequately addressed on the Davis-Besse docket in NRC InspectionReport 50-346/03-19, LERs 2003-002 and 2002-005, and FENOC lettersdated February 27, 2004 (ML040620456), November 26, 2003(ML033370836), and October 24, 2003 (ML040890175). In summary,Davis-Besse had met the requirements of NRC Generic Letter 98-04 and hadcommitted to maintain the Nuclear Safety-Related Protective CoatingsProgram for coating material controls and coating application to structuresand components located within the Containment.
 
==Enclosure==
L-1 1-238Page 96 of 98In 2006, the Nuclear Safety-Related Protective Coatings Programdocumented inspection findings in the Corrective Action Program for theCycle 14 refueling outage. Inspection findings were:" Epoxy topcoat cracking and peeling areas observed on severalembedded plates on east and north surfaces of the east Once-ThroughSteam Generator (OTSG) enclosure (D-ring) walls. Approximately50 square feet of coating material was cracked or peeling. The coatingwas applied during initial plant construction." Upper edge of the west D-ring at edge for the missile shield supportshelf had approximately one square foot of peeled coating. Thebaseplate for a pipe whip restraint located on the east D-ring hadapproximately two square feet of peeled material.* Approximately one square foot of degraded material was observed onan embedded plate (approximate elevation 625'-0") for west staircaserestraint and on two pipe restraint baseplates at an elevation ofapproximately 650'-0".Corrective actions taken were to add the quantity of failed protective coatingmaterial to the Non-DBA Qualified Coating Inventory and to plan removal andrework of the failed coating material.In 2008, NRC Integrated Inspection Report 05000346/2008-03 described theimplementation of the Davis-Besse actions documented in theFebruary 28, 2008, response to Generic Letter 2004-02, "Potential Impact ofDebris Blockage on Emergency Recirculation during Design Basis Accidentsat Pressurized Water Reactors." The Davis-Besse resolution of Generic Letter2004-02 included the installation of a significantly larger strainer withincontainment. The debris source term was also significantly reduced throughremoval of nearly all fibrous insulation and completely stripping and recoatingthe containment dome. Detailed analyses that used bounding limits for debrisgeneration, transport and head loss effect were performed using theNEI 04-07, "Pressurized Water Reactor Sump Performance EvaluationMethodology," and associated NRC Safety Evaluation Report (SER)methods, with permitted deviations. The NRC inspectors reviewed theengineering change packages (ECPs) associated with modifications installed,procedure changes and programmatic controls implemented, and changes forthe Updated Safety Analysis Report (USAR) in response to GenericLetter 2004-02. No findings of significance were identified.In 2008, the Nuclear Safety-Related Protective Coatings Programdocumented inspection findings in the Corrective Action Program for the
 
==Enclosure==
L-1 1-238Page 97 of 98Cycle 15 refueling outage. General coating conditions in Containmentremained acceptable. Inspection findings were:" Blistering of containment dome coating material in two locations. Thedegraded material was quantified and added to the Non-DBA QualifiedProtective Coatings Inventory.* Peeling of containment vessel top coat material behind the polar craneaccess ladder between elevations 714' to 722'. The degraded coatingmaterial was removed." Rusting of containment penetrations P3, P4, P5, P6, P7, P9, P10and P11 was identified and evaluated.* Peeling of epoxy top coat on bottom of northeast, upper OTSG 1-1support." Flaking paint on a hot leg platform brace adjacent to the OTSG wasquantified and added to the Non-DBA Qualified Protective CoatingsInventory." Peeled top coat material was found on a lower snubber mounting forOTSG 1-2.Several areas of degradation which were noted during this outage hadpreviously been identified and are to be reworked. The degraded material inthese areas has been included in the Non-DBA Qualified ProtectiveCoatings Inventory.In 2011, the Nuclear Safety-Related Protective Coatings Programdocumented inspection findings in the Corrective Action Program for theCycle 16 refueling outage. General coating conditions in Containmentremained good. Inspection findings were:* Blistering of containment vessel coating material in two locationsadjacent to the polar crane access ladder at approximately the660' elevation. The degraded material has been removed." Peeling coating material on structural steel for the elevation 610'-0" hotleg platform. The degraded material has been removed." Rusting of containment penetrations identified and previouslyevaluated. Rework of these penetrations is currently planned to beperformed per work order during the Cycle 18 refueling outagescheduled for spring 2014.
 
==Enclosure==
L-1 1-238Page 98 of 98Peeling of epoxy top coat on bottom of northeast, upper OTSG 1-1support. The degraded material was quantified and added to theNon-DBA Qualified Protective Coatings Inventory." Flaking paint on hot leg platform brace adjacent to the OTSG. Thedegraded material was quantified and added to the Non-DBA QualifiedProtective Coatings Inventory." Peeled top coat material was found on a lower snubber mounting forOTSG 1-2. The degrade material was quantified and added to theNon-DBA Qualified Protective Coatings Inventory.Several areas of degradation which were noted during this outage had beenidentified previously and are currently planned to be reworked during theCycle 18 refueling outage. The degraded material in those areas has beenincluded in the Non-DBA Qualified Protective Coatings Inventory.Enhancements:None.ConclusionThe Nuclear Safety-Related Protective Coatings Program is an existingprogram that has been demonstrated to be capable of monitoring theperformance of coatings inside containment. Proper maintenance ofprotective coatings has ensured that the quantities of unqualified anddegraded qualified coatings inside containment are maintained below theacceptance limits. The continued implementation of the Nuclear Safety-Related Protective Coatings Program provides reasonable assurance thatthe effects of aging will be managed such that the Service Level 1 protectivecoatings and other coatings in containment are maintained consistent withthe current licensing basis for the period of extended operation.
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