ML12094A383: Difference between revisions
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{{#Wiki_filter:FENOCFirstEnergy Nuclear Operating Company5501 North State Route 2Oak Harbor. Ohio 43449Barry S. AllenVice President -Nuclear419-321-7676Fax: 419-321-7582March 9, 2012L-12-01510 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), LicenseRenewal Application Amendment No. 24, and Revised License Renewal ApplicationBoundary DrawingsBy 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 letter dated December 27, 2011(ML 1333A396), the Nuclear Regulatory Commission (NRC) requested additionalinformation to complete its review of the License Renewal Application (LRA).The NRC letter contained four requests for additional information (RAIs). The FENOCresponse to one of the four RAI responses, RAI 3.1.2.2.16-3, was submitted to the NRCby letter dated January 13, 2011 (ML12018A338).Attachment 1 provides the FENOC responses to two of the four RAIs (B.1.4-2 andB.1.4-3) in the NRC letter. The submittal date for these responses was extendedfollowing discussion with Mr. Samuel Cuadrado de Jesus, NRC Project Manager,because additional time was needed for development of the responses due tocoordination with multiple site and Fleet departments. Attachment 1 also providessupplemental information on the topics listed below. The NRC request is shown in boldtext followed by the FENOC response:* RAI 2.1-3 regarding abandoned equipment;* RAI 3.1.2.2-2 regarding Reactor Vessel Internals aging management; and* LRA Section 4.2, "Reactor Vessel Neutron Embrittlement." az,It-f1I "H,1-1- ý- | |||
Davis-Besse Nuclear Power Station, Unit No. 1L-12-015Page 2Enclosure A provides Amendment No. 24 to the DBNPS LRA. Enclosure B providesnew and revised License Renewal Boundary Drawings.The FENOC response to the fourth of four RAIs (B.2.39-13) in the NRC letter is plannedto be provided to the NRC by March 30, 2012, due to the need to evaluate andincorporate information from the Davis-Besse Shield Building concrete cracking RootCause Analysis Report that was recently completed and submitted to the NRC byFENOC letter dated February 27, 2012 (ML1 20600056).Attachment 2 identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)in this document. Any other actions discussed in the submittal represent intended orplanned actions by FENOC; they are described only as information and are notRegulatory Commitments. Please notify Mr. Clifford I. Custer, Project Manager- FleetLicense Renewal, at (724) 682-7139 of any questions regarding this document orassociated Regulatory Commitments.I declare under penalty of perjury that the foregoing is true and correct. Executed onMarch 1¶ , 2012.Sincerely,Barry S. AllenAttachment:1. Reply to Requests for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,Sections B.1.4, 2.1, 3.1.2 and 4.22. Regulatory Commitment ListEnclosures:A. Amendment No. 24 to the DBNPS License Renewal ApplicationB. New and Revised DBNPS License Renewal Application Boundary Drawingscc: NRC DLR Project ManagerNRC Region III Administrator Davis-Besse Nuclear Power Station, Unit No. 1L-12-015Page 3cc: w/o Attachment or EnclosuresNRC DLR DirectorNRR DORL Project ManagerNRC Resident InspectorUtility Radiological Safety Board | |||
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1L-12-015Reply to Requests for Additional Information for the Review of the.Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,Sections B.1.4, 2.1, 3.1.2 and 4.2Page 1 of 24Section B.1.4Question RAI B.1.4-2Background:In request for additional information (RAI) B.1.4-1, issued on May 19, 2011, thestaff asked the applicant to describe the programmatic activities that will be usedto continually identify aging issues, evaluate them, and as necessary, enhance theaging management programs (AMPs) or develop new AMPs for license renewal. Inits response dated June 24, 2011, the applicant stated that it currently has aprocedurally controlled operating experience review process, as required byNUREG-0737, "Clarification of TMI Action Plan Requirements," Item I.C.5,"Procedures for Feedback of Operating Experience to Plant Staff." The applicantstated that this process provides for the systematic identification and transfer oflessons learned from site and industry experience into-fleet and station processesto prevent events and enhance the safety'and reliability of its operations.Issue:The applicant's response provided a general description of how it considersoperating experience on an ongoing basis; however, it does not directly addressseveral areas in RAI B.1.4-1 on which the staff requested information. Further, theapplicant's response did not provide specific information on how the operatingexperience review activities address issues specific to aging. The staff identifiedthe following issues with the applicant's response:(a) The applicant did not describe the sources of plant-specific operatingexperience information that it monitors on an ongoing basis. Additionaldetails are needed to determine whether the applicant will consider anadequate scope of information from which to identify potential operatingexperience related to aging.(b) It is not clear whether the applicant only reviews certain sources foroperating experience information. Additional information is needed todetermine whether the applicant's processes would preclude theconsideration of relevant operating experience Information, because it isnot from a prescribed source. | |||
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1L-12-015Page 2 of 24(c) The staff requested that the applicant indicate which guidance documentsrequire monitoring. Theapplicant did not indicate whether it considersguidance documents to be a source of operating experience information.(d) The applicant did not describe its criteria for identifying and categorizingoperating experience items as related to aging.(e) The staff requested that the applicant describe training provided to plantpersonnel. The applicant did not describe training on aging issues, nor didit indicate whether the training will be provided for those plant personnelresponsible for screening, assigning, evaluating, and submitting operatingexperience items.(f) The applicant did not describe how evaluations of operating experiencerelated to aging consider the potentially affected plant* systems, structures, and components,4 materials,* environments,* aging effects,* aging mechanisms, and" AMPs.(g) The applicant did not describe how it will consider as operating experiencethe results of the inspections, tests, analyses, etc., conducted throughimplementation of the AMPs.(h) The applicant did not describe the records of operating experienceevaluations or how it retains those records.(i) The applicant stated that operating experience evaluations are prioritizedwith due dates procedurally specified based on the potential significanceof the issue; however, the applicant did not provide details on theevaluation schedules or how it determines the relative significance ofissues. It is therefore unclear whether the operating experienceevaluations will be completed in a timely manner or whether they will be-appropriately prioritized.(j) The applicant stated that it enters operating experience that potentiallyrepresents~a condition adverse to quality into the corrective actionprogram; however, the applicant did not explain how a "condition adverseto quality" includes aging. Additional information, is needed to determine | |||
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1L-12-015Page 3 of 24whether the corrective action program has a threshold appropriate tocapture items related to aging.(k) The applicant did not describe criteria for considering when AMPs shouldbe modified or new AMPs developed due to operating experience. It alsodid not describe how it implements these kinds of changes or how itensures the changes are implemented in a timely manner.(I) The applicant stated that it shares lessons learned with other utilities topromote industry-wide safety and reliability; however, the applicant did notprovide criteria for reporting its plant-specific operating experience onage-related degradation to the industry.Request:Provide a response to each item below.(a) Describe the sources of plant-specific operating experience that aremonitored on an ongoing basis to identify potential aging issues.(b) Indicate whether plant-specific and industry operating experience is onlyconsidered from a prescribed list of sources. If only prescribed sources areconsidered, provide ajustification as to why it is unnecessary to considerother sources.(c) Indicate whether guidance documents are considered as a source ofoperating experience information. If they are considered as a potentialsource, provide a plan for considering the content of guidancedocuments, such as the GALL Report,, as operating experience applicableto aging management.(d) Describe how operating experience issues will be identified andcategorized as related to aging.(e) Describe the training requirements on aging issues for those plantpersonnel responsible for screening, evaluating, and submitting operatingexperience items.(f) Describe how evaluations of operating experience issues related to agingwill consider the following:* systems, structures, or componentsa materials* environments0 aging effect | |||
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1L-12-015Page 4 of 24* aging mechanisms* AMPs(g) Describe how the results of the AMP inspections, tests, analyses, etc., willbe considered as operating experience.(h) Describe the operating experience evaluation records with respect to whatis considered for aging. Indicate whether-these records are maintained inauditable and retrievable form.(i) Provide details on the operating experience evaluation schedules andjustify why they provide for timely evaluations. Also, describe how therelative significance of operating experience items is determined so thatthe reviews can be prioritized appropriately.(j) Justify why the corrective action program has an appropriate threshold forcapturing issues concerning aging.(k) Describe the criteria for considering when AMPs should be modified ornew AMPs developed due to operating experience. Also, describe theprocess for implementing changes to the AMPs or for implementing newAMPs; describe how these changes are implemented in a timely manner.(I) Provide criteria for reporting plant-specific operating experience onage-related degradation to the industry.If enhancements are necessary, provide an implementation schedule forincorporating them into the existing programmatic operating experiencereview activities.RESPONSE RAI B.1.4-2(a) Describe the sources of plant-specific operating experience that are monitoredon an ongoing basis to identify potential aging issues.The sources of plant-specific operating experience to identify potential aging issuesinclude the following:-- FENOC Corrective Action Program -adverse conditions are documented inthe Corrective Action Program, including, when appropriate, the cause andactions necessary to correct and/or prevent recurrence. Adverse conditionsas defined in the Corrective Action Program include any event, defect,characteristic, state or activity that prohibits or detracts from safe, efficientnuclear plant operation or a condition that could credibly impact nuclearsafety, personnel safety, plant reliability or non-compliance with federal, state, | |||
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1L-12-015Page 5 of 24or local regulations. Adverse conditions that are failures, malfunctions,deficiencies, deviations, defective hardware and non-conformances, orhuman performance, programmatic, organizational, or managementweaknesses that adversely affect Quality (Q), Augmented Quality (AQ), ornuclear safety-related equipment, programs, or processes, are consideredconditions adverse to quality. Adverse conditions include conditions adverseto quality, plant reliability issues, any concern that should be trended (e.g.,personnel contamination, personnel safety,,and unexpected plant equipmentfailures), and conditions that have significance within a regulatory context.FENOC does not differentiate whether an adverse condition is aging relatedwhen initiating a condition report; adverse conditions resulting from agingrelated issues are .documented in the Corrective Action Program in the samemanner as any other adverse condition. The Corrective Action Program,therefore, is considered the primary source of plant-specific operatingexperience since adverse conditions, including aging-related adverseconditions, are documented in the program database." Plant maintenance activities -maintenance activities -such as preventivemaintenance tasks, inspections, examinations, surveillances, tests oranalyses, are sources of plant-specific operating experience, includingpotential aging issues. Adverse conditions identified during the performanceof these types of activities are documented in the Corrective Action Program.* Plant Operator tours -Operator tours are performed routinely throughout thestation and provide the opportunity to identify adverse system, structure orcomponent conditions, including aging issues. Adverse conditions identifiedduring the performance of Plant Operator tours are documented in theCorrective Action Program." System Engineer walkdowns -Periodic System Engineer walkdownsperformed as part of the System Performance Monitoring Program providethe opportunity to identify adverse system, structure or componentconditions, including aging issues. Adverse conditions identified during thecourse of System Engineer walkdowns are documented in the CorrectiveAction Program.* Aging Management Program activities -preventive maintenance tasks,inspections or examinations performed at the direction of aging managementprograms are sources of plant-specific operating experience, including agingissues. Additionally, trending of program inspection or examination resultsand effectiveness reviews required by select aging management programsmay also provide a source of plant-specific operating experience, includingaging issues. Adverse conditions identified during the performance of thesetypes of activities are documented in the Corrective Action Program. | |||
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1L-12-015Page 6 of 24-(b) Indicate whether plant-specific and industry operating experience is onlyconsidered from a prescribed list of sources. 'If only- prescribed sources areconsidered, provide a justification as to why it is unnecessary to considerother sources.FENOC does not consider plant-specific and industry operating experience onlyfrom a prescribed list of sources.For plant-specific operating experience, FENOC considers input from many sources,such as those described in the response to question (a), above. Adverse conditionsidentified from these sources are documented in the FENOC Corrective ActionProgram. The Corrective Action Program procedure does not include a prescribedlist of sources from which to identify adverse conditions; rather, the procedureprovides examples of potential adverse conditions so that plant personnel willidentify plant issues from a wide range of sources.Although the Operating Experience Program does not restrict the use of othersources of operating experience, it does require- screening of industry operatingexperience from a list of prescribed sources at a minimum. The list of prescribedoperating experience sources includes:* Institute of Nuclear Power Operations (INPO) Event Reports;* INPO Significant Operating Experience Reports (SOERs);* INPO Operating Experience Reports;* Nuclear Regulatory Commission (NRC) Information Notices (INs);* NRC Regulatory Issue Summaries (RISs); and,* NRC Regulatory changes.The Operating Experience Program procedure also allows for processing operatingexperience from additional sources not listed above on a case-by-case basis.Other programs and processes, such as those provided in the examples identifiedbelow, are also available to FENOC for obtaining industry operating experience andaging-related information:The Continuous Equipment Performance Improvement process (deriveddirectly from INPO's AP-913, "Equipment Reliability Process Description")requires periodic updates to the process based on evaluation of plant-specificand industry operating experience from sources such as new vendorrecommendations, Electric Power Research Institute (EPRI) preventivemaintenance templates, the INPO Equipment Performance and InformationExchange (EPIX) system, or aging studies. | |||
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1L-12-015Page 7 of 24The Materials Degradation Management Program is an industry initiative toadopt Nuclear Energy Institute (NEI) guidance for managementof materialissues. The Program is applicable to materials of construction within theReactor Coolant System Pressure Boundary, although-it can be applied tomaterials of construction within other systems. The Materials DegradationManagement Program ensures, among other issues, that materialS operatingexperience is shared among utilities.Industry owners groups, vendors, and EPRI committees, industrymanaged groups and Users groups share information that can help identifyaging concerns.Industry benchmarking lessons learned.(c) Indicate whether guidance documents are considered as a source of operatingexperience information. If they are considered as a potential source, provide aplan for considering the content of guidance documents, such as the GALLReport, as operating experience applicable to aging management.FENOC considers NRC Information Notices, Regulatory Issue Summaries, andRegulatory changes as sources of operating experience. However, FENOC does notconsider NRC guidance documents as sources of operating experience information.Guidance documents, such as the NRC NUREG-Series publications, are reports orbrochures on regulatory decisions, results of research, results of incidentinvestigations, and other technical and administrative information. These documentsare not sources of operating experience; however, the NRC may use NUREGs tocompile and report the results of operating experience research from industrysources. For example, new License Renewal Interim Staff Guidance documents orrevisions to NUREG-1800, "Standard Review Plan for Review of License RenewalApplications for Nuclear Power Plants," and NUREG-1 801, "Generic Aging LessonsLearned (GALL) Report," provide, in addition to license renewal guidance, the resultsof research on previously-identified industry operating experience and the associatedlessons-learned, and are historical documents by the time they are issued.FENOC is actively involved with numerous industry users groups and committees,which regularly share information on new industry operating experience. Changesto guidance documents are often identified and discussed at these industrymeetings, allowing FENOC to be informed of the issuance of new or revisedguidance documents.While the Operating Experience Program procedure does not include NRC guidancedocuments on the list of operating experience sources to be reviewed, the proceduredoes include Regulatory Issue Summaries, which can be used by the NRC to inform | |||
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1L-12-015Page 8 of 24licensees of changes to guidance documents. For example, FENOC was madeaware of Revision 2 to NUREG-1800 and 1801 through the Nuclear Energy Institute(NEI) License Renewal Working Groups and through Regulatory Issue Summary2011-05, "Information on Revision 2 to the Generic Aging Lessons Learned Reportfor License Renewal of Nuclear Power Plants." Regulatory Issue Summary 2011-05was screened by FENOC using the Operating Experience Program.(d) Describe how operating experience issues will be identified and categorizedas related to aging.Plant-specific aging-related operating experience issues documented in the FENOCCorrective Action Program are processed and investigated in the same manner asother adverse conditions. However, the Corrective Action Program NuclearOperating Procedure is planned to be revised to require that a condition reportinvestigation of an aging-related issue for structures and passive componentsinclude consideration of the affected structure or component, material, environment,aging effect, aging mechanism, and aging management program, with feedback tothe affected aging management program owner for consideration of the impact toaging management program effectiveness.Industry aging-related operating experience items are received and screened nodifferently than other industry operating experience items received through existingoperating experience pipelines. Aging-related operating experience items are nottypically received with an "aging" flag or designator; rather, the operatingexperience items identify a condition or event experienced at another station. Forexample, corrosion-related or cracking failures may not necessarily be designatedas "aging-related."The Fleet Operating Experience Program Manager and Station OperatingExperience Coordinators review INPO's Nuclear Network and incomingcorrespondence for new external operating experience items. A screening process isperformed for the external operating experience items to determine the susceptibilityof a similar and unacceptable event or condition occurring at FENOC. If theoperating experience item is screened as not applicable (NA) to FENOC, no furtheraction is taken. An operating experience item that is determined to be applicable toFENOC is further screened as follows:If the operating experience item represents a potential, operability orreportability concern, the operating crew Shift Manager or Shift Engineer isnotified and a condition report is'written to document the issue in theCorrective Action Program.* If the operating experience item represents an adverse condition, a conditionreport is written to document the issue in the Corrective Action Program. | |||
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1L-12-015Page 9 of 24If the operating experience item is identified for an "Evaluation RequiredReview," a formal evaluation with supervisory or management oversightis initiated.If the operating experience item requires further screening, it is submitted toan Operating Experience Coordinator peer group, which may include subjectmatter experts as needed, for a more detailed review to determinesusceptibility to FENOC stations.If the operating experience item is screened as applicable, but may notrequire a document evaluation, the item is considered "Information Only," andis-listed in the Weekly Operating Experience Summary.o. The Section Operating Experience Coordinator is responsible forreviewing the Weekly Operating Experience Summary and determiningwhether any of the items require an evaluation, enlisting subject matterexperts as necessary to assist in the determination.** These Information Only operating experience reports are alsocommunicated and distributed to personnel via the Weekly OperatingExperience Summary. No specific evaluation is required for operatingexperience items identified as Information Only reviews. However, ifthe reviewer determines that further evaluation is required, thereviewer should initiate an Evaluation Required Review by contactingthe Fleet or Station Operating Experience Coordinator.The Operating Experience Program Nuclear Operating Business Practice is plannedto be revised to require that an Evaluation Required Review of aging-relatedoperating experience issues for structures and passive components includeconsideration of the affected structure or component, material, environment, agingeffect, aging mechanism, and aging management program, with feedback to theaffected aging management program owner for consideration of the impact to agingmanagement program effectiveness.Additionally, the Corrective Action Review Board Nuclear Operating BusinessPractice is planned to be revised to ensure that the Corrective Action Review Boardquestions whether aging was considered for condition report investigations that arereviewed by the Board.See the enhancements descriptions at the end of this response. | |||
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1L-12-015Page 10 of 24(e) Describe the training requirements on aging issues for those plantpersonnel responsible for screening, evaluating, and submitting operatingexperience items.The training requirements for personnel responsible for screening, evaluating andsubmitting (to the industry) operating experience items consist of completion ofposition-specific document and procedure reviews, proficiency demonstrations andinterviews that provide personnel with the ability to independently perform theactivity. Mentoring is used during the training process to provide direction, coaching,and oversight by a recognized technical expert who is qualified in the specific taskrequirements to ensure job performance requirements are understood andcompetency is achieved.A training "needs analysis" is planned to determine and document recommendedenhancements to the training requirements for those personnel responsible forscreening, evaluating and submitting (to the industry) aging-related operatingexperience items.See the enhancements descriptions at the end of this response.(f) Describe how evaluations of operating experience issues related to aging willconsider the following:* systems, structures, or components* materials* environments* aging effect* aging mechanisms* AMPsFor plant-specific aging-related operating experience issues, the Corrective ActionProgram Nuclear Operating Procedure is planned to be revised to require that acondition report investigation of an aging-related issue for structures and passivecomponents include consideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging management program, withfeedback to the affected aging management program owner for consideration of theimpact to aging management program effectiveness.For industry aging-related operating experience items, the Operating ExperienceProgram Nuclear Operating Business Practice is planned to be revised to requirethat an Evaluation Required Review of aging-related operating experience issues for | |||
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1L-12-015Page 11 of 24structures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and agingmanagement program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(g) Describe how the results of the AMP inspections, tests, analyses, etc., will beconsidered as operating experience.Degraded or non-conforming conditions or results that do not meet the definedacceptance criteria identified during aging management program inspections, tests,analyses or other program activities are considered adverse conditions and aredocumented in condition reports using the FENOC Corrective Action Program.These results and the associated condition report are considered operatingexperience for the affected aging management program.The applicable site or fleet Management Review Board is responsible to reviewcondition reports to identify internal operating experience that has the potential to beshared with the other FENOC sites and the industry. The following attributes areused to make the determination:* Important to nuclear, public, radiological, and personnel safety* Important to power generation capability* Events with important generic implications* Events for which an investigation was performed and the lessons learnedwould be beneficial to know about if the event had occurred at another station* Events required to be reported by INPO Event Report (IER) Level 1 orSOER recommendationsThe results of aging management program inspections, tests, analyses or otherprogram activities that meet the defined acceptance criteria are also consideredoperating experience for the affected aging management program. These results areused as feedback to the aging management program for trending purposes, and forevaluation to determine, based on the component, material, environment and agingeffect combinations managed, whether the frequency of future inspections needs tobe adjusted or new inspections established.The Corrective Action Program Nuclear Operating Procedure is planned to berevised to require that a condition report investigation of an aging-related issue for,structures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and aging | |||
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1L-12-015Page 12 of 24management program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(h) Describe the operating experience evaluation records with respect to what isconsidered for aging. Indicate whether these records are maintained inauditable and retrievable form.Operating experience evaluation records are processed using existing plantprocedures. Plant-specific operating experience is documented in condition reportsand processed using the FENOC Corrective Action Program. Aging issues thatwould be captured in the Corrective Action Program include adverse conditionssuch as the following:0 aging-related degradation that adversely affects or threatens plant systems,structures or components;* adverse aging-related trends identified in system health reports or agingmanagement programs;0 aging management program weaknesses or programs identified asineffective; or,* results that do not meet the acceptance criteria from inspections,examinations, tests, analyses, or other activities performed under an agingmanagement program.Condition reports for adverse conditions and related documents captured inthe Corrective Action Program database are quality records and are auditableand retrievable.In a similar manner, industry operating experience sources are monitored andoperating experience, including aging-related operating experience, is entered intothe -Operating Experience Program database. External operating experience itemsare exposed to a screening process that provides for three possible outcomes:, Not applicable (NA) to FENOC, and no further action is taken;* Applicable-to FENOC, but may not require a documented evaluation, and aretherefore considered "Information Only;" or,* Applicable-to FENOC and require an evaluation ("Evaluation Required Review").Operating experience reports that are screened as "Information Only" are unlikely toaffect nuclear or personnel safety, plant reliability or availability, or are unlikely toneed follow-up actions. Information Only operating experience items are. | |||
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1L-12-015Page 13 of 24communicated and distributed to personnel in a weekly operating experiencesummary. If a reviewer later determines that further evaluation is required, thereviewer should initiate an "Evaluation Required Review" by contacting the Fleet orStation Operating Experience Coordinator.An Evaluation Required Review is performed for operating experience reports thatare determined to require a FENOC or site evaluation, or when the initial screeningidentifies the need for further evaluation and/or actions to address the operatingexperience issue. The degree of rigor involved in evaluating operating experienceinformation should be based on the significance of the issue identified. Documentscaptured in the Operating Experience Program database are retrievable.(i) Provide details on the operating experience evaluation schedules and justifywhy they provide for timely evaluations. Also, describe how the relativesignificance of operating experience items is determined so that the reviewscan be prioritized appropriately.Plant-specific operating experience is documented in condition reports andprocessed using the FENOC Corrective Action Program. Supervisor reviews ofcondition reports are expected to be performed as soon as possible, normally withinone business day of initiation of the condition report. The Management ReviewBoard, which consists of a collegial review and concurrence by the Managers from arepresentative cross-section of plant disciplines, reviews new condition reports dailyand establishes the due dates for condition report evaluations. Due dates are setbased on the safety significance of the issue; the standard due date for conditionreport evaluations is 30 days. For plant-specific operating experience that should beshared with the industry, the goal is to issue the report to the industry within 50 daysof the origination date of the associated condition report.Industry operating experience is documented in the Operating Experience databaseand processed using the FENOC Operating Experience Program. If the operatingexperience issue is screened such that an Evaluation Required Review is assignedand no operability, reportability or adverse conditions are identified, then time isavailable to properly evaluate the issue. For operating experience Evaluation RequiredReviews for INPO Level 1 and 2 Event Reports, INPO is to be informed of the FENOCaction plan within 150 days. For Evaluation Required Reviews for other operatingexperience items, the evaluation due date is initially established as 150 days.The determinations of the relative significance and the resulting determination oftimeliness for the evaluation of operating experience items are subjective, but arebased on the plant knowledge and experience of the reviewers. Considerationsinclude items such as the real or potential impact to nuclear safety, generation risk,plant-operation, systems, structures or components, programs, potential operability | |||
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1L-12-015Page 14 of 24or reportability concerns, or whether an adverse condition exists. The real orpotential impacts may not be obvious, and research may need to be performed todetermine whether the station has similar systems, structures or components,materials, or conditions such that the station would be affected.(j) Justify why the corrective action program has an appropriate threshold forcapturing issues concerning aging.The FENOC Corrective Action Program has an appropriate threshold for capturingissues concerning aging because the program requires that adverse conditions,regardless of their nature (including aging-related degradation), are documented in acondition report in the Corrective Action Program. The training provided to plantpersonnel coupled with procedural guidance and supervisory and managementoversight ensures that adverse conditions, including issues concerning aging, areappropriately captured and evaluated. Periodic regulatory and utility-basedassessments and effectiveness reviews of the Corrective Action Program are usedto confirm that adverse conditions are appropriately addressed.(k) Describe the criteria for considering when AMPs should be modified or newAMPs developed due to operating experience. Also, describe the process forimplementing changes to the AMPs or for implementing new AMPs; describehow these changes are implemented in a timely manner.The FENOC Corrective Action Program requires development of actions that areeffective at addressing conditions adverse to quality. Examples of effectivecorrection actions that address aging may include modification of existing ordevelopment of new aging management programs. Adverse aging conditions maybe identified via plant-specific or industry operating experience that indicate one ormore of the following:* a new aging effect is identified;" the applicable aging effect or aging mechanism for a given combination ofsystem, structure or component, material, environment, aging effect andaging mechanism is not being effectively managed;" adverse aging-related trends are identified in system health reports or agingmanagement programs;" aging management program weaknesses are identified or programs areidentified as ineffective; or, | |||
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1L-12-015Page 15 of 24, results are obtained that do not meet the acceptance criteria from inspections,examinations, tests, analyses, or other activities performed under an agingmanagement program.Revision of existing or development of new aging management programs based onoperating experience evaluations would be performed through corrective actionsusing the Corrective Action Program, or by action items identified in the OperatingExperience Program database. Assigned program owners would develop revisionsto their assigned aging management program implementing procedures based onthe results of an evaluation of the operating experience issues. A regulatoryapplicability review and, as applicable, a 10 CFR 50.!59 review are performed toconfirm whether the change can be made without prior NRC approval, and whetherthe change impacts any information in the Davis-Besse Updated Safety AnalysisReport. New aging management program implementing procedures would bedeveloped based on the activities involved and the systems, structures orcomponents affected. The revised or new aging management program implementingprocedure would undergo an internal FENOC review and approval process. The duedates for corrective actions and Operating Experience Program action, items areclosely tracked and managed.For plant-specific aging-related operating experience issues, the Corrective ActionProgram Nuclear Operating Procedure is planned to be revised to require that acondition report investigation of an aging-related issue for structures and passivecomponents include consideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging management prograrfi, withfeedback to the affected aging management program owner for consideration of theimpact to aging management program effectiveness.For industry aging-related operating experience items, the Operating ExperienceProgram Nuclear Operating Business Practice is planned to be revised to requirethat an Evaluation Required Review of aging-related operating experience issues forstructures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and, agingmanagement program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(I) Provide criteria for reporting plant-specific operating experience onage-related degradation to the industry.Noteworthy plant-specific operating experience is shared with the other FENOCsites and the industry. The applicable site or fleet Management Review Board is | |||
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1L-12-015Page 16 of 24responsible to identify internal operating experience that has the potential to beshared with the other FENOC sites and the industry. The following attributes areused to make the determination:0 Important to nuclear, public, radiological, and personnel safety* Important to power generation capability* Events with important generic implications* Events for which an investigation was performed and the lessons learnedwould be beneficial to know about if the event had occurred at another station* Events required to be reported by INPO Event Report (IER) Level 1 orSOER recommendationsThe Operating Experience Program procedure provides the guidance on sharinginternal operating experience.If enhancements are necessary, provide an implementation schedule forincorporating them into the existing programmatic operating experience reviewactivities.The following training activities and procedure changes will be completed on or beforeDecember 31, 2012:1. Perform a training needs'analysis to determine and-document recommendedenhancements to the training requirements for those plant personnel responsiblefor screening, evaluating and submitting (to the industry) aging-related operatingexperience items. Based on the results of the training needs analysis, identify theappropriate training materials.2. Revise Nuclear Operating Business Practice NOBP-LP-2100, "FENOCOperating Experience Process," to require that an Evaluation Required Reviewof aging-related operating experience issues for structures and passivecomponents includes:a. consideration of the material, environment, aging effect, agingmechanism, and aging management program for the affected structure orcomponent; and,b. a provision for feedback to the affected aging management programowner for consideration of the impact to aging management programeffectiveness. | |||
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1L-12-015Page 17 of 243. Revise Nuclear Operating Procedure NOP-LP-2001, "Corrective ActionProgram," to require that a condition report investigation of an aging-related issuefor structures and passive components includes:a. consideration of the material, environment, aging effect, agingmechanism, and aging management program for the affected structure orcomponent; and,b. a provision for feedback to the affected aging management programowner for consideration of the impact to aging management programeffectiveness.4. Revise Nuclear Operating Business Practice NOBP-LP-2008, "FENOCCorrective Action Review Board," to ensure that the Corrective Action ReviewBoard questions whether aging was considered for condition report investigationsthat are reviewed by the Board.See Attachment 2 to this letter for the regulatory commitments.Question RAI B.1.4-3Background:In RAI B.1.4-1, the staff asked the applicant to provide, in accordance with10 CFR 54.21(d), a USAR supplement a summary description of theprogrammatic activities for the ongoing review of operating experience, asrequired by 10 CFR 54.21(d). By letter dated August 17, 2011, the applicantprovided this description:Existing FENOC processes require reviews of relevant site andindustry operating experience and periodic benchmarking toensure program enhancements are identified and implemented.Such ongoing reviews identify potential needs for agingmanagement program revisions to ensure their effectivenessthroughout the period of extended operation.Issue:As described above in RAI B.1.4-2, the applicant described generally how itintends to consider operating experience on an ongoing basis; however, it didnot provide specific information on how its operating experience review activitiesaddress issues related to aging. Similarly, the above entry for USAR supplement | |||
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1.L-12-015Page 18 of 24also lacks details on how aging is considered in the ongoing operatingexperience reviews.Request:,Consistent with the response to RAI B.1.4-2, provide additional details in theUSAR supplement on how the ongoing operating experience review activitiesaddress issues specific to aging.RESPONSE RAI B.1.4-3FENOC replaces the operating experience summary description in License RenewalApplication (LRA) Section A. 1, "Summary Descriptions of Aging Management Programsand Activities," provided by letter dated August 17, 2011 (ML1 1231A966), with a newdescription that is consistent with the FENOC response to RAI B.1.4-2, above.See Enclosure A to this letter for the revision to the DBNPS LRA.Section 2.1Supplemental Question RAI 2.1-3 Abandoned EquipmentIn a supplemental response to RAI 2.1-3 titled, "SUPPLEMENTAL RESPONSE -Abandoned Equipment," submitted by letter dated September 16, 2011(ML11264A059), FENOC provided a plan to ensure abandoned equipment isidentified, isolated and drained, as follows:1. Determine the scope of abandoned equipment -includes review of Piping& Instrumentation Diagrams (P&IDs), plant walkdowns, and review of theShift Operations Management System (eSOMS) clearance database.2. Determine the status of abandoned equipment -includes review of systemstatus files and the eSOMS database for as-left valve positions, walkdownsto validate valve position status, and ultrasonic testing to confirm thatabandoned piping is drained.3. Place abandoned equipment in a configuration that will not impactsafety-related equipment -create and implement Operations EvolutionOrders to isolate and drain abandoned systems with fluids, and create and | |||
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1L-12-015Page 19 of 24implement Document*Change Requests as necessary to correct theconfiguration of the plant as shown on plant drawings.SUPPLEMENTAL RESPONSE RAI 2.1-3 Abandoned EquipmentFENOC completed the actions to ensure abandoned equipment is identified, isolatedand drained as provided in the plan submitted by FENOC letter datedSeptember 16, 2011 (ML11264A059). Identification of the scope of abandonedequipment that could impact safety-related equipment was determined through a reviewof Piping & Instrumentation Diagrams (P&IDs), plant walkdowns, and review of the ShiftOperations Management System (eSOMS) clearance database.The status of the abandoned equipment that could impact safety-related equipment wasdetermined through review of system status files and the eSOMS database for as-leftvalve positions, walkdowns to validate valve position status, and ultrasonic testing toconfirm that abandoned piping is drained.Abandoned equipment that could impact safety-related equipment was verified to beisolated and drained with the exception of components associated with the ServiceWater System intake crib air bubbler compressors, and the Miscellaneous LiquidRadwaste System degasifier skid, miscellaneous waste evaporator skid, evaporatorstorage tank pumps, and primary water transfer pumps. The subject components areadded tothe scope of license renewal in accordance with 10 CFR 54.4(a)(2), as follows:* The Lake Erie elevation is such that it provides a nonisolable source of water tothe discharge piping associated with the abandoned intake crib air bubblercompressors. The components associated with these discharge lines containfluid, and are, therefore, added to the scope of license renewal. License renewalboundary drawing LR-M012E is revised to highlight these additional components.LRA Section 2.3.3.26, "Service Water System," is revised to add license renewalboundary drawing LR-M012E to the list of license renewal boundary drawings thatdepict the evaluation boundaries of the Service Water System. LRA Table 2.3.3-26, "Service Water System Components Subject to Aging Management Review,"is revised to identify a structural integrity function for the orifice component type.LRA Table 3.3.2-26, "Aging Management Review Results -Service WaterSystem," is revised to provide the updated aging management review results forthese additional components.The components associated with the abandoned degasifier skid, miscellaneouswaste evaporator skid, evaporator storage tank pumps, and primary watertransfer pumps contain fluid or are not sufficiently isolated to remain drained, andare, therefore, added to the scope of license renewal. New License renewalboundary drawings LR-M01OD Sheet 2, LR-M036C Sheet 2 and LR-M039BSheet 2 were created, and license renewal boundary drawings LR-M01OD | |||
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1L-12-015Page 20 of 24Sheet 1, LR-M020B, LR-M036C Sheet 1, LR-M037C, LR-M037D, LR-M037E,LR-M037F, LR-M039A, LR-M039B Sheet 1 and LR-M042B are revised tohighlight the additional components. LRA Section 2.3.3.21, "Miscellaneous LiquidRadwaste System," is revised to add additional license renewal boundarydrawings to the list of boundary drawings that depict the evaluation boundaries ofthe system. LRA Table 2.3.3-21, "Miscellaneous Liquid Radwaste SystemComponents Subject to Aging Management Review," is revised to include thenew component types. LRA Table 3.3.2-21, "Aging Management Review Results-Miscellaneous Liquid Radwaste System," is revised to provide the updatedaging management review results for the additional components.See Enclosure A to this letter for the revision to the DBNPS LRA.See Enclosure B to this letter for the revision to the LRA Boundary Drawings.Section 3.1.2Supplemental Question RAI 3.1.2.2-2 Reactor Vessel Internals Aging ManagementThe NRC initiated a telephone conference call with FENOC on January 24, 2012,to discuss the pressurized water reactor (PWR) Reactor Vessel Internals Programand how FENOC plans to implement MRP-227-A, "Materials Reliability Program:Pressurized Water Reactor Internals Inspection and Evaluation Guidelines(MRP-227-A)." The staff wanted to know FENOC's plans to address anydifferences between MRP-227 and MRP-227-A.FENOC responded that a review of MRP-227-A dated December 2011 against theprevious version of MRP-227 dated December 2008 was in progress to identifyany changes needed to the Davis-Besse PWR Reactor Vessel Internals (RVI)Program or to the aging management review (AMR) results provided in LicenseRenewal Application (LRA) Table 3.1.2-2, "Aging Management Review Results -Reactor Vessel Internals." FENOC noted that, at a minimum, several changes arerequired to LRA Table 3.1.2-2.FENOC agreed to provide a response to address the differences betweenMRP-227-A and the previous version of MRP-227, and provide the necessaryupdates for the reactor vessel internals AMR results and the PWR Reactor VesselInternals (RVI) Program. | |||
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1L-12-015Page 21 of 24The staff requested that FENOC be very clear as to what'has and what has notchanged, and to identify what was reviewed.SUPPLEMENTAL RESPONSE RAI 3.1.2.2-2 Reactor Vessel Internals Aging ManagementAging Management Review Results for the Reactor Vessel InternalsIn response to RAI 3.1.2.2-3 submitted by FENOC letter dated September 16,-2011(ML1 1264A059), LRA Table 3.1.2-2, "Aging Management Review Results -ReactorVessel Internals," was replaced in its entirety. These aging management review (AMR)results were based on Electric Power Research Institute (EPRI) Report 1016596,"Materials Reliability Program: Pressurized Water Reactor Internals Inspection andEvaluation Guidelines (MRP-227-Rev. 0)," dated December 2008. This report has sincebeen updated with the changes proposed by the NRC in the safety evaluation forMRP-227 as well as changes proposed by the EPRI Materials Reliability Program inresponse to NRC requests for additional information (RAIs) and issued as EPRI Report1022863, "Materials Reliability Program: Pressurized Water Reactor InternalsInspection and Evaluation Guidelines (MRP-227-A)," dated December 2011.FENOC completed a review of MRP-227-A dated December 2011 and NRC safetyevaluation, "Revision 1 to the Final Safety Evaluation of Electric Power Research Institute(EPRI Report), Materials Reliability Program (MRP) Report 1016596 (MRP-227),Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and EvaluationGuidelines" (TAC No. ME0680)," against the previous version of MRP-227 datedDecember 2008, and determined that changes are required to the AMR results providedin LRA Table 3.1.2-2. A summary of the required table changes are listed as follows:Row 4 -control rod guide tube (CRGT) spacer casting changed from an expansioncomponent with primary component link of core support shield (CSS) cast outletnozzles, CSS vent valve discs or in-core monitoring instrumentation (IMI) guidetube spiders to a primary component with no expansion components. This revisedclassification is consistent with Section 3.3.7 of the NRC safety evaluation.Row 10 -CSS cast outlet nozzles changed from a primary component withexpansion component of CRGT spacer casting to a 'no additional measures'component. This revised classification is consistent with Table 3-1 ofMRP-227-A. As provided in note 2 of the table, "Thermal Embrittlement (TE)revised from 'P' to 'A' based on review of material data (ferrite content) fromONS-3 and DB; final grouping accordingly changed from 'Primary' to 'NoAdditional Measures."'"Row 12 -As discussed in Section' 3.7 of the-safety evaluation, the CSS ventvalve disc was determined to be an active component and not subject to agingmanagement. Therefore, row 12 is changed to "Not used." | |||
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1L-12-015Page 22 of 24" Row 13 -As discussed in Section 3.7 of the safety evaluation, the CSS ventvalve disc shaft was determined to be an active component and not subject toaging management. Therefore, row 13 is changed to "Not used."* Row 20 -In addition to the aging effects of cracking due to irradiation-assistedstress corrosion cracking (IASCC) and reduction in fracture toughness, the agingeffects of cracking due to fatigue, loss of material and loss of preload were addedfor the core barrel-to-former (CBF) bolts. These revised aging effects ateconsistent with Table 3-1 of MRP-227-A.* Row 21 -In addition to the aging effects of cracking due to IASCC and reductionin fracture toughness, the aging effects of cracking due to fatigue, loss of materialand loss of preload were added for the baffle-to-former (FB) bolts. These revisedaging effects are consistent with Table 3-1 of MRP-227-A.* Row 22 -In addition to the aging effects of cracking due to fatigue and reductionin fracture toughness, the aging effects of loss of material and loss of preloadwere added for the baffle-to-baffle (BB) bolts -internal. These revised agingeffects are consistent with Table 3-1 of MRP-227-A." Row 23 -In addition to the aging effects of cracking due to fatigue, cracking dueto IASCC and reduction in fracture toughness, the aging effects of loss of materialand loss of preload were added for the baffle-to-baffle (BB) bolts -external.These revised aging effects are consistent with Table 3-1 of MRP-227-A.* Row 42 -Since the CRGT spacer casting was changed to a primary component(see Row 4 above), it is deleted as an expansion component for the IMI guide tubespiders. This revised classification is consistent with Table 4-1 of MRP-227-A." Row 43 -Since the CRGT spacer casting was changed to a primary component(see Row 4 above), it is deleted as an expansion component for the IMI guidetube spider-to-lower grid rib section welds. This revised classification isconsistent with Table 4-1 of MRP-227-A." Plant-specific note 0114 -This note addressed the classification of the flowdistributor (FD) bolts and their locking devices but is no longer needed for thatpurpose since Table 4-1 of MRP-227-A now shows the component as a primarycomponent with expansion components of upper thermal shield (UTS) bolts andtheir locking devices, lower thermal shield (LTS) bolts and their locking devices,and surveillance specimen holder tube (SSHT) bolts and their locking devices.This revised classification is consistent with Section 4.1.3 of the NRC safetyevaluation. Plant-specific note 0114 is now used to provide clarification thatcomponents assigned to the no additional measures group were determined-tobe below the screening criteria for the applicable degradation mechanisms, orwere classified under this category due to Failure Modes, Effects, and Criticality | |||
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1L-12-015Page 23 of 24Analyses (FMECA) and functionality analysis findings, and therefore, no furtheraction is required by MRP-227-A for managing the aging of these components.PWR Reactor Vessel Internals ProgramIn response to RAI B.2.32-1 submitted by FENOC letter dated September 16, 2011(ML1 1264A059), the Davis-Besse PWR Reactor Vessel Internals Program wasreplaced in its entirety. This revised program, provided in LRA Section B.2.32 and thecorresponding Updated Safety Analysis Report (USAR) Supplement (LRA SectionA.1.32), addressed each of the five plant-specific aging management programinformation requirements identified in Section 3.5.1 of the NRC safety evaluation, "FinalSafety Evaluation of EPRI Report, Materials Reliability Program Report 1016596(MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection andEvaluation Guidelines" (TAC No. ME0680)," dated June 22, 2011.FENOC has completed a review of the five plant-specific aging management programinformation requirements identified in Section 3.5.1 of the NRC safety evaluation,"Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRIReport), Materials Reliability Program (MRP)_Report 1016596 (MRP-227), Revision 0,"Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines"(TAC No. ME0680)," dated December 16, 2011, and determined that no changes arerequired to the Davis-Besse PWR Reactor Vessel Internals Program relative to addressingeach of the five plant-specific aging management program information requirements.As noted in the above discussion (see "Row 13" bulleted item) that addressed therequired changes to LRA Table 3.1.2-2, the CSS vent valve disc shaft was determinedto be an active component and not subject to aging management. Therefore, LRASections A.1.32 and B.2.32, both titled "PWR Reactor Vessel Internals Program," arerevised to delete any discussion of the vent valve disc shaft.In addition, the 5th paragraph of the "Detection of Aging Effects" program element forLRA Section B.2.32 was provided to address the flow distributor (FD) bolts classificationand examination method/coverage/frequency. Since MRP-227-A has incorporated thechanges outlined in Section 4.1.3 of the safety evaluation relative to the flow distributor(FD) bolts, the 5th paragraph is no longer needed and is therefore deleted.Also, LRA Sections A.1.32 and B.2.32 are also revised to incorporate MRP-227-Aversus MRP-227, Revision 0, as the program implementation document.See Enclosure A to this letter for the revision to the DBNPS LRA. | |||
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1L-12-015Page 24 of 24Section 4.2Supplemental Question Section 4.2 Reactor Vessel Neutron EmbrittlementThe NRC initiated a telephone conference call with FENOC on February 9, 2012, todiscuss time-limited aging analyses (TLAAs) associated with Section 4.2 of theLicense Renewal Application (LRA). Two of the topics discussed during thetelephone conference are as follows.In LRA Section 4.2.1, the neutron fluence analysis for 52 EFPY (60-years ofoperation) was dispositioned as not a time-limited aging analysis (TLAA). Thestaff's position is that neutron fluence is a TLAA. FENOC agreed to change theLRA to disposition the neutron fluence analysis as a TLAA using10 CFR 54.21(c)(1)(ii).In LRA Section 4.2.6 including the associated USAR Supplement Section A.2.2.6,discussion was provided to address the future replacement of the reactor vesselclosure head. FENOC agreed to change the LRA to acknowledge that the reactorvessel closure head was replaced in the fall of 2011.SUPPLEMENTAL RESPONSE Section 4.2 Reactor Vessel Neutron EmbrittlementAs provided in LRA Section 4.2.1, "Neutron Fluence," a neutron fluence analysis validfor 52 effective full power years (EFPY) has been prepared for the reactor vesselbeltline materials to bound the projected value of 50.3 EFPY for 60-years of operation.LRA Section 4.2.1, including the associated USAR Supplement Section A.2.2.1,"Neutron Fluence," and LRA Table 4.1-1, "Time-Limited Aging Analyses," are revised todisposition the neutron fluence analysis as a time-limited aging analysis (TLAA) inaccordance with 10 CFR 54.21 (c)(1)(ii).LRA Section 4.2.6, "intergranular Separation (Underclad Cracking)," including theassociated USAR Supplement Section A.2.2.6, "lntergranular Separation -UndercladCracking," addressed the future replacement of the reactor vessel closure head/headflange schedule for the fall of 2011. Also, the operating experience program element forLRA Section B.2.29, "Nickel-Alloy Reactor Vessel Closure Head Nozzles Program,"addressed the future replacement of the head. The reactor vessel closure head/headflange was replaced in the fall of year 2011. Therefore, LRA Sections 4.2.6, A.2.2.6 andB.2.29 are revised to provide this updated status for the reactor vessel closure head.See Enclosure A to this letter for the revision to the DBNPS LRA. | |||
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2L-12-015Regulatory Commitment ListPage 1 of 2The following list identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)in this document. Any other actions discussed in the submittal represent intended orplanned actions by FENOC; they are described only as information and are notRegulatory Commitments. Please notify Mr. Clifford I. Custer, Project Manager -FleetLicense Renewal, at (724) 682-7139 of any questions regarding this document orassociated Regulatory Commitments.Regulatory CommitmentDue Date1. Perform a training needs analysis to determine December 31, 2012and document recommended enhancements tothe training requirements for those plantpersonnel responsible for screening, evaluatingand submitting (to the industry) aging-relatedoperating experience items. Based on the resultsof the training needs analysis, identify theappropriate training materials.2. Revise Nuclear Operating Business Practice December 31, 2012NOBP-LP-2100, "FENOC Operating ExperienceProcess," to require that an Evaluation RequiredReview of aging-related operating experienceissues for structures and passive componentsincludes:a. consideration of the material, environment,aging effect, aging mechanism, and agingmanagement program for the affectedstructure or component; and,b. a provision for feedback to the affectedaging management program owner forconsideration of the impact to agingmanagement program effectiveness. | |||
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2L-12-015Page 2 of 2Regulatory CommitmentDue Date-I-3. Revise Nuclear Operating ProcedureNOP-LP-2001, "Corrective Action Program," torequire that a condition report investigation of anaging-related issue for structures and passivecomponents includes:a. consideration of the material, environment,aging effect, aging mechanism, and agingmanagement program for the affectedstructure or component; and,b. a provision for feedback to the affectedaging management program owner forconsideration of the impact to agingmanagement program effectiveness.December 31, 2012 "4. Revise Nuclear Operating Business Practice December 31, 2012NOBP-LP-2008, "FENOC Corrective ActionReview Board," to ensure that the CorrectiveAction Review Board-questions whether agingwas considered for condition report investigationsthat are reviewed by the Board. | |||
==Enclosure== | |||
ADavis-Besse Nuclear Power Station, Unit No. I (DBNPS)Letter L-12-015Amendment No. 24 to theDBNPS License Renewal ArplicationPage 1 of 51License Renewal ApplicationSections AffectedSection 2.3.3.21Table 2.3.3-21Section 2.3.3.26Table 2.3.3-26Table 3.1.2-2Table 3.1 Plant-Specific NotesTable 3.3.2-21Table .3.3.2-26Section 4.2.1.3Section 4.2.6Section A. 1Section A.1.32Section A.2.2.1Section A.2.2.6Section B.2.29Section B.2.32Table 4.1-1The 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 fined-ou and added text underlined. | |||
==Enclosure== | |||
AL-12-015Page 2 of 51Affected LRA Section LRA Page No. Affected Paragraph and Sentence2.3.3.21 Page 2.3-113 License Renewal Drawings,9 additional drawingsrIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Section2.3.3.21, "Miscellaneous Liquid Radwaste System," is revised to add nineadditional license renewal boundary drawings to the list of "License RenewalDrawings," to read as follows:License Renewal DrawingsThe following license renewal drawings depict the evaluation boundaries for thesystem components within the scope of license renewal:LR-MOIOD Sheet 1, LR-MOIOD Sheet 2, LR-MO2OB, LR-M031A, LR-M033A,LR-M036A, LR-M036C Sheet 1, LR-M036C Sheet 2, LR-M037C, LR-M037D,LR-M037E, LR-M037F, LR-M037G, LR-M039A, LR-M039B Sheet 1,LR-M039B Sheet 2, LR-M042B, LR-M045, LR-M046, LR-M281N13 | |||
==Enclosure== | |||
AL-12-015Page 3 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceTable 2.3.3-21 Page 2.3-114 12 new rowsIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Table2.3.3-21, "Miscellaneous Liquid Radwaste System Components Subject to AgingManagement Review," is revised to include 12 new rows for new componenttypes, and reads as follows:Component Type Intended Function(as defined in Table 2.0-1)Heat exchanger (shell) -Degasifier Heat Exchangers Structural integrityHeat exchanger (shell) -Waste Evaporator Heat, Structural integritExchangersPump casing -Anti Foam Tank Pump (DB-P55 WM) Structural integrityPump casing -Degasifier Discharge Pump Structural integrityPump casing -Evaporator Storage Tank Pumps Structural integrity(DB-P53-1 & 2)Pump casing -Primary Water Transfer Pumps Structural integrity(DB-P41-1, 2)Pump casing -Waste Evaporator Bottoms Pump Structural-integrityPump casing -Waste. Evaporator Distillate Pump Structural integrityPump casing -Waste Evaporator Vacuum Pump Structural integrityTank -Anti Foam Tank (DB-T30) Structural integrityTank -Degasifier (DB-S3) Structural integrityTank -Waste Evaporator Concentrator (DB-S2) Structural integrity | |||
==Enclosure== | |||
AL-12-015Page 4 of 51Affected LRA SectionLRA Page No.Page 2.3-128Affected Paragraph and SentenceLicense Renewal Drawings,1 additional drawing2.3.3.26In response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Section2.3.3.26, "Service Water System," is revised to add one additional license renewalboundary drawing to the list of "License Renewal Drawings," to read as follows:License Renewal DrawingsThe following license renewal drawings depict the evaluation boundaries for thesystem components within the scope of license renewal:LR-MOO6D, LR-MO12E, LR-M036A, LR-M036B, LR-M041A, LR-M041B, LR-M041CAffected LRA SectionLRA Page No.Page 2.3-130Affected Paragraph and SentenceTable 2.3.3-26Intended FunctionIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Table2.3.3-26, "Service Water System Components Subject to Aging ManagementReview," is revised to include the "structural integrity" intended function for theorifice component type, and reads as follows:Component Type Intended Function(as defined in Table 2.0-1)Pressure boundaryOrifice Structural integrityThrottling. | |||
==Enclosure== | |||
AL-12-015Page 5 of 51Affected LRA SectionTable 3.1.2-2Table 3.1.2 Plant-Specific NotesLRA Page No.Pages 3.1-60thru 3.1-121Page 3.1-187Affected Paragraph and Sentence27 Rows, various columnsNote 0114In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals Aging Management, 27 rows of LRATable 3.1.2-2, "Aging Management Review Results -Reactor Vessel Internals," and Plant-Specific Note 0114,previously replaced in their entirety by FENOC letter dated September 16, 2011 (ML1 1264A059), are revisedto read as follows:MIo0y A- IOU uuweIs-tL- PWR ReactorPlenum Cover Borated Reactor Vessel Internals None ABottomNickel Coolant with Cracking -SCC (IV.B4.RP- None 0Flange Welds Support Alloy 0114(no additional Neutron Fluence PWR Water 236)measures component) Chemistry | |||
==Enclosure== | |||
AL-12-015Page 6 of 514(expansion componentwith p~mriry Gomponentlink of CSS Cast OutletNozz~es, CSS VentValve~ DiBscs orIM!(primary componentwith no expansioncomponents)SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracturetoughnessPWR ReactorVessel InternalsIV. B4-4(IV.B4.RP-242)3.1.1-80ACRGT Rod GuideTubes Stainless Borated Reactor Loss of PWR Reactor None A5 Support Coolant with Loss NoneSupot olat it material -wear Vessel Internals 23V) B41P14on(no additional Steel Neutron Fluence 237) 0114measures component)CRGT Rod GuideSectors Stainless Borated Reactor Loss of PWR Reactor None A(n diinlSupport Sanes Coolant with(I.4R- Nn6 Suppona Steel Neutron Fluence material -wear Vessel Internals (IV.B4.RP- None 0114meaddition ent)237)___measures component) _________________________________________ | |||
==Enclosure== | |||
AL-12-015Page 7 of 51Upper Core Barrel(UCB) Bolts (originalbolts) and their lockingdevicesBolt: Cracking -SCCPWVK ReactorVessel InternalsIV.1B4-20(IV.B4.RP-248)3.1.1-37A0444PWR WaterChemistry8(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,.and SSHT Bolts andtheir locking devices)SupportStainlessSteelBorated ReactorCoolant withNeutron FluenceLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV. B4-01(IV.B4.RP-243)C3.1.1-22____________________ _________________ ___________________________ __________________________ +/- ___________________________ I _____________________ L _______________ I __________ | |||
==Enclosure== | |||
AL-12-015Page 8 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRo~wNo.9Component TypeIntendedFunction(s)MaterialEnvironmentAging EffectRequiringManagementAging 'ManagementProgramNUREG-1801Rev. I Item(Rev. 2 Item)Table IItemNotesF t I F I t I IUpper' Core Barrel(UCB) Bolts(replacement bolts) andtheir locking devices(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)CSS Cast OutletNozzles(primar-y cOMPonentcomponents 414 fWlOW,:CRGT Spacer Casting)(no additionalmeasures component)CSS Vent Valve Discs(primary ,OMPnentwith e ,PWR ReactorVessel InternalsBolt: Cracking -SCCIV.1B4-20(IV.B4.RP-248)3.1.1-37A0444SupportBolt:NickelAlloyLockingDevices:StainlessSteelBorated ReactorCoolant withNeutron FluenceBolt:Cumulative IV. B4-37 3 1.1-05 Afatigue damage (IV. B4. R-53)-fatiguePWR WaterChemistryLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV. B4-01(IV.B4.RP-243)3.1.1-22C10SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracture.toughnessPWR ReactorVessel InternalsIV. B4-21(IV.B4.RP-253)3.1.1-80A011412Naomt Us ed Not Used&49petF46W~GR*nGO;astAustenoStainlessStee4Berated Reacte,GeelaR*-tNeutron FluenceReduetien infraetU~etoughnessPKVFR ReactorVessRel InternalI-V -84-24(~L84Rpz344. 0AI I_____ I ____________________ I __________ 4 ___________________________ 1 __________________________ 4 ... ______________ | |||
==Enclosure== | |||
AL-12-015Page 9 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type ntn Material Environment Requiring Management Rev. I Item NotesManagement Program (Rev. 2 Item)CSS Vent Valv-e DiscShaf tB4 1...ompoen Berated Reacto Re& in PW .I-3 4.RPQ13 ,M no e n suppe--o4 Vessel , ,,; 344 2 AGOwit noexpnsio L)p# Neutron Fluenc toughness OV-.B4.RP-compenents) Not Used \Core Barre[AssemblPWR ReactorAlloy X-750 Vessel Internals None. ACore Barrel-to-Former Borated Reactor Cracking -SCC (IV.B4.RP- None 011415 Plate Dowel Support Nickel PWR Water 236)15 PaeDwlSpot Alloy Coolant with Chemistry(n o additional Neutron Fluence Reduction in Nonemeasures component) fracture PWR Reactor (IV.B4.RP- None AVessel Internals. " 0114toughness 237)Alloy X-750 PWR ReactorDowel-to-Core Barrel Borated Reactor Vessel Internals None A16 Cylinder Fillet Welds Support Nickel Coolant with Cracking -SCC (IV.B4.RP- None(no additional Alloy Neutron Fluence PWR Water 236) 0114measures component) ChemistryThermal Shield Upper Cracking -Restraint Cap Borated ReactoIr ueNone.17 Screws (Not Exposed) Support Stess Coolant with Loss of (IV.B4.RP- None(no additional Neutron Fluence material -wear Vessel Internals 237) 0114measures component) Loss of preload | |||
==Enclosure== | |||
AL-12-015Page 10 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1o. Component Material Environment Requiring Management Rev. 1 Item NotesManagement Program (Rev. 2 Item) ItemIV.B4-07(IV.B4.RP-PWR Reactor 244);Cracking -Vessel Internals (IV.B4.RP-Core Barrel-to-Former IASCC, Fatigue 238) 3.1.1-30 A(CBF) Bolts Borated Reactor PWR Water None20 (expansion component Support Stainless Coolant with Chemistry (IV.B4.RP-with primary component Steel Neutron Fluencelink of FB Bolts) Reduction infracture IV.1B4-01toughness PWR Reactor (IV.B4.RP-Loss of Vessel Internals 243); 3.1.1-22 Amaterial -wear (IV. B4. R P-239)Loss of preload | |||
==Enclosure== | |||
AL-12-015Page 11 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging_ NUREG-1801 Table INo. Component Type Intended Material Environment Requiring Management Rev. I Item NotesFunction(s) Management Program (Rev. 2 Item) ItemIV.B4-07PWR Reactor (IV.B4.RP-Cracking -Vessel Internals 241) 3.1.1-30 ABaffle-to-Former (FB) IASCC, FatiQue PWR Water None 3Bolts___Borated Reactor Chemistry (IV.B4.RP-21 (primary component Support Stainless Colntwt 37L5121 with expansion Steel Coolant withcomponents of BB Neutron Fluence fractureBolts and CBF Bolts) toughness PWR Reactor IV.B4-01Loss of Vessel Internals (IV.B4.RP- 3.1.1-22 Amaterial -wear 240)Loss of preloadPWR ReactorVessel Internals NoneCracking -(IV.B4.RP- None ABaffle-to-Baffle (BB) fatigue PWR Water 375)Bolts -internal Borated Reactor Chemistry22 (expansion component Support Stainless Coolant with Reduction inwith primary component Steel Neutron Fluence fracturelink of FB Bolts) toughness PWR Reactor IV. B4-01Loss of Vessel Internals (IV.B4.RP- 3.1.1-22 Amaterial -wear 243)Loss of preload | |||
==Enclosure== | |||
AL-12-015Page' 12 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow .Intended Aging Effect Aging NUREG-1801 Table 1Row Component Type Intended Material Environment Requiring Management Rev. I Item NotesNo. Function(s) Management Program (Rev. 2 Item) ItemPWR ReactorCracking -Vessel Internals Noneratinge (IV.B4.RP- None Afatigue PWR Water 375)ChemistryBaffle-to-Baffle (RB) PWR Reactor IV.1B4-07Bolts -external Boae Vessel Internals (IV.B4.RP-Stainless Borated Reactor Cracking- 244); 3.1.1-30 A23 (expansion component Support Steel Coolant with IASCC PWR Water (IV.B4.RP-with primary component Neutron Fluence Chemistry 238)link of FB Bolts) Reduction infracture IV. B4-01toughness PWR Reactor (IV.B4.RP-Loss of Vessel Internals 243);material -wear (IV.B4.RP-Loss of preload 239)Lower Core Barrel PWR Reactor(LCB) Bolts (original) Bolt: Cracking -Vessel Internals IV.R4-13 Aand their locking SCC (IV.B4.RP-. 3.1.1-37 0-44devices PWR Water 247)(primary component Borated Reactor Chemistry26 with expansion Support Ste Coolant withcomponents of UTS Steel Neutron Fluence LockingBolts and their locking Device: PWR Reactor IV. B4-01devices, LTS Bolts and Loss of Vessel Internals (IV.B4.RPJ 3.1.1-22 Ctheir locking devices, material -wear 243)and SSHT Bolts andtheir locking devices) | |||
==Enclosure== | |||
AL-12-015Page 13 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type Function(s) Material Environment Requiring Management Rev. I Item NotesNo. Fnis _______ Management Program (Rev. 2 Item) Item27Lower Core Barrel(LCB) Bolts(replacement) and theirlocking devices(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)Bolt: Cracking -SCCPWR ReactorVessel InternalsIV.B4-13(IV.B4.RP-247)3.1.1-37SupportBolt:NickelAlloyLockingDevices:StainlessSteelBorated ReactorCoolant withNeutron FluenceBolt:Cumulative .IV.B4-37 3.1.1-05 Afatigue damage (IV.B4.R-53)-fatiguePWR WaterChemistryA0444LockingDevice:.Loss ofmaterial -wearPWR ReactorVessel InternalsIV.1B4-01(IV.B4.RP-243)3.1.1-22C30Alloy X-750 Dowel-to-Upper Grid Rib SectionBottom Flange Welds(noadditionalmeasures component)SupportNickelAlloyBorated ReactorCoolant withNeutron FluenceCracking -SCCPWR ReactorVessel InternalsPWR WaterChemistryNone(IV.B4.RP-236)NoneA011435Alloy X-750 Dowel-to-Lower GridShell Forging Welds(no additionalmeasures component)SupportNickelAlloyBorated ReactorCoolant withNeutron FluenceCracking -SCCPWR ReactorVessel InternalsNone(IV.B4.RP-236)NoneA0114PWR WaterChemistry | |||
==Enclosure== | |||
AL-12-015Page 14 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item NotesManagement Program (Rev. 2 Item)PWR ReactorVessel Internals NoneAAlloy X-750 Dowel-to- Cracking -ALower Grid Rib Borated Reactor SCC, IASCC (IV.B4.RP- NoneNickel PWR Water 236) 011436 Section Welds Support Alloy Coolant withChemistry(no additional Neutron Fluence Chemistrymeasures component) Reduction in PWR Reactor None Afracture Vessel Internals (IV.B4.RP- None 0114toughness 237)Lower Grid Rib-to-Shell Cracking -Forging Cap Borated Reactor fatigue NoneScrews Stainless PWR Reactor f37 Support Steel Coolant with Loss of Vessel Internals o (IV.B4.RP- None 0114(no additional Neutron Fluence material -wear 237) Noe-11measures component) Loss of preloadLower Grid Support Cracking -Post Pipe Cap Borated Reactor fatigue None38 Screws Support Stainless Coolant with Loss of V esl ter (IV.B4.RP- None A(no additional Steel Neutron Fluence material -wear 237)measures component) Loss of preload | |||
==Enclosure== | |||
AL-12-015Page 15 of 51ti-ow uisirnutor (I-u)Bolts and their lockingdevicesPWi< ieactorVessel InternalsBolt: CrackingSCCIV.B4-25(IV.B4.RP-256)A ,3.1.1-37 A044-440(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)SupportStainlessSteelPWR WaterChemistryBorated ReactorCoolant withNeutron FluenceLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV.B4-01(IV.B4.RP-243)3.1.1-22 1 CAlloy X-750,Dowel-to- PWR ReactorFlow Distributor Borated Reactor Vessel Internals None41 Flange Welds Support Alloy Coolant with Cracking -SCC (IV.B4.RP- None 0(no additional .Neutron Fluence PWR Water 236) 0114measures component) Chemistry | |||
==Enclosure== | |||
AL-12-015Page 16 of 5142IMI (5uiae I uve spicters(primary componentwith expansioncomponents of-GRGTSpacor- Casting an~Lower Fuel AssemblySupport Pads: Pad,Pad-to-Rib SectionWeld, Cap Screw andassociated LockingWeld, Alloy X-750Dowel and Alloy X-750Dowel Locking Weld)SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracturetoughnessPWR ReactorVessel InternalsIV. B4-28(IV.B4.RP-258)3.1.1-80A_____________ ____________ J. __________________ -I _________________ .1 __________________ .1. ______________ 1 _________ ~I. _______ | |||
==Enclosure== | |||
AL-12-015Page 17 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table INo. Component Type Fntinds) Material Environment Requiring Management Rev. I Item NotesNo. Function(s) Management Program (Rev. 2 Item) "ItemIMI Guide Tube Spider-to-Lower Grid RibSection Welds(primary componentwith expansioncomponents of-CRGTSpa.er. Casting an Stainless Borated Reactor Reduction in PWR Reactor IV.1B4-3143 Lower Fuel Assembly Support Steel Coolant with fracture Vessel Internals (IV.B4.RP- 3.1-.1-22 ASupport Pads: Pad, Neutron Fluence toughness 259)Pad-to-Rib SectionWeld, Cap Screw andassociated LockingWeld, Alloy X- 750Dowel and Alloy X-750Dowel Locking Weld)Plant-Specific Notes:'0114 Flow distribto ~F~blswr esined as a primary component per- MRP--227-, Rev. 0 as amended by the safet evaluation.Due to thiS haG , t--FDbolts are not anepnso omponent for- the UCB and LOB bolts.Components assigned to the "No Additional Measures" group were determined to be below the screening criteria for theaiplicable degradation mechanisms, or were classified under this category due to Failure Modes, Effects, and Criticality Analyses(FMECA) and functionality analysis findings, and therefore, no further action is required by MRP-227-A for managing the aging ofthese components. | |||
==Enclosure== | |||
AL-12-015Page 18 of 51Affected LRA SectionLRA Page No..Page 3.3-411thru 3.3-417Affected Paragraph and SentenceTable 3.3.2-2136 New Rows; and,Rows 55, 60, 65, 70, 75, 79, & 84, "Notes" columnIn response to Supplemental RAI 2.1-3 Abandoned Equipment, 36 new rows are added to LRATable 3.3.2-21, "Aging Management Review Results -Miscellaneous Liquid Radwaste System." Additionally,during the review, it was identified that generic Note "A" was assigned in the "Notes" column for rows 55, 60,65, 70, 75, 79, and 84, but should have been Note "C". LRA Table 3.3.2-21 is revised to include new rows andcorrect assigned notes, as follows:Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemTAging Effect NUREG-Row Component Intended Material Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItemHeatExchanger Collection, Drainage,(shell) -Structural Stainless Raw water Loss of and Treatment VII.C15 3.3.1-79 E-Degasifier integrity Steel (Internal) material Components VHeat InspectionExchangersHeatExchanger Air with(shell) -Structural Stainless borated water None None VIIJ-16 3.3.1-99 C-Degasifier inte~grity Steel leakageHeat (External)Exchangers | |||
==Enclosure== | |||
AL-12-015Page 19 of 51Table 3.3.2-21 Aging Management Review Results -.Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table NotesNo. Type Function(s) Program Volume 2 itemManagement Item .IHeatExchanger Air-indoor(shell) -] Structural Stainless Aridoesher itrity steel uncontrolled None None VII.J-15 3.3.1-94 C-- Detasfir Steel (External)HeatExchangersHeatExchanger(shaell) -Collection, Drainage,(shell)_- Structural Stainless Raw water Loss of and Treatment-- Waste ..Sel (nenl aeilCmoet VII.CI-15 3.3.1-79 EEvaporator integrit Steel (Internal) material ComponenHeat InspectionExchangersHeatExchanger. Air with(shell) Structural Stainless borated waterWaste None None VII. J- 16 3.3.1-99 CEvaporator inert Steel leakage .Nn oe___ ___Heat o (External)Exchangers | |||
==Enclosure== | |||
AL-12-015Page 20 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 NotesNo. Type Function(s) Program Volume 2 ItemManagement I ItemHeatExchanger(shell) Structural Stainless Air-indoorWaste uncontrolled None None VII.J-15 3.3.1-94 C.Evaporator i Steel (External)HeatExchangersPump casing Collection, Drainage.-Anti Foam Structural Stainless Raw water Loss of and Treatment-- Tank Pump VII._ C1-5 _.31-9_TnB- integrity Steel (Internal) material Components VICI-15 3.3.1-79 EP55 WM) InspectionPump casing Air with-Anti Foam ArwtTank Fuoa Structural Stainless borated water-- Tank Pump itgiy Sel eageNone None VII.J-16 3.3.1-99 A(DB- Steel leakaeP55 WM) (External)Pump casing-Anti Foam Air-indooStructural Stainless Air-indoorTank Pump Integrity Steel uncontrolled None None VII.J-15 3.3.1-94 A(DB- (External)P55 WM) | |||
==Enclosure== | |||
AL-12-0.15Page 21 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemdAging Effect NUREG-Row Component Intended Material Environment Requiring Agn aaeet 10, Tbe1 NotesNo. Type Function(s) M Program Volume 2 ItemMaeral Enirnmn Rqurig gigManagement18, Tal NoeItemPump casing Collection, Drainage,-Degasifier Structural Stainless Raw water Loss of and Treatment VIICI-15 3.3.1-79 E-Dischare integrity Steel (Internal) material ComponentsPUMP InspectionPump casing Air with-Degasifier Structural Stainless borated water None None VIIJ-16 3.3.1-99 ADischarge integrit Steel leakagePump (External)Pump casing Air-indoor-Degasifier Structural Stainless Aridouncontrolled None None VII.J-15 3.3.1-94 ADischarge integrity Steel (External)PumpPump Casing Collection, Drainage,-Evaporator-- Storage Tank Structural Stainless Raw water Loss of and Treatment VII.C1-15 3.3.1-79 EPumPs (DB- integrit Steel (Internal) material Components,P53-1,. 2) InspectionPump Casing Air with-Evaporato Structural Stainless borated water-- Storage Tank Strut Stail boate None None VII. J- 16 3.3.1-99 APumps (DB- i Steel leakageP53-1, 2) (External) | |||
==Enclosure== | |||
AL-12-015Page 22 of 51Table 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 ItemItemPump Casing-EvaporaTa Structural Stainless Air-indoor-Storage Tank Strur Stanle uncontrolled None None VII.J-15 3.3.1-94 APumps (DB- n Steel (ExternalP53-1, 2)Pump casing Collection, Drainage,-Primary Structural Stainless Raw water Loss of and Treatment-- Water VI1_C1-15 3.3.1-79Transfer inte-grit Steel (Internal) material Componentsu.Trane Inspection* PumpPump casing Air with-Primary Structural Stainless borated water-- Water SNone None VII.J-16 3.3.1-99 ATransfer ,negt Steel leakagePum.n (External)Pump casing-Primary Structural Stainless Air-indoorWater uncontrolled None None VII.J-15 3.3.1-94 ATransfer integrity Steel (External)PumPPump casing Collection, Drainage,-Waste Structural Stainless Raw water Loss of and Treatment VEvaporator integrit Steel (Internal) material Components VII-IS 3.3.1-79 EBottoms InspectionPump I I____ | |||
==Enclosure== | |||
AL-12-015Page 23 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAgin EfectNUREG-Row Component Aging Management 1801, Table tggfcNMaterial Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItemPump casing Air with-Waste Structural Stainless borated water-- Evaporator integrit Steel leakage None None VII.J-16 3.3.1-99 ABottoms (External)'PumPPump casing-Waste. Air-indoo-- EWasorator Structural Stainless Air-indoor-- EvaporatorS uncontrolled None None VII.J-15 3.3.1-94 ABottoms n Steel (External)PumpPump casin Collection, Drainage,-Waste Structural Stainless Raw water " Loss of and Treatment-E rato integrit Steel (Internal) material ComponentsDistillate InspetoPump IpectionPump casing Air with-Waste Structural Stainless borated water None None VII.J-16 3.3.1-99 ADEvsporator integrity Steel leakageNDistillate (External)Pump casing-Waste Structural Stainless Air-indoor-- Evaporator i t Steel uncontrolled None None VII.J-15 3.3.1-94 ADistillate (External)Pump_ | |||
==Enclosure== | |||
AL-12-015Page 24 of 51Table 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 Management Program Volume 2 ItemItemPump casingPm Wasig Collection, Drainage,Structural Stainless Raw water Loss of and TreatmentEvaporator integrity Steel (Internal) material Components VII.C1-15 3.3.1-79 EVacuum IsetoPumpPump casing Air with-Waste Structural Stainless borated waterEvaporator inte.grit Steel leakage None None VII.J-16 3.3.1-99 AVacuum inertSte leagPumum (External)Pump casing-Waste Structural Stainless Air-indoorEvaporator integrit Steel uncontrolled None None VII.J-15 3.3.1-94 A.Vacuum (External)PumpTank -Anti Collection, Drainage,TankoamAntik Structural Stainless Raw water Loss of and TreatmentFoam Tank integrity Steel (Internal) material Components VII.C-15 3.3.1-79 E(DB-T30) InspectionSTank -Anti Air withFoam Tank Structural Stainless borated water None None VII.J-16 3.3.1-99 C(DBT30) integrity Steel leaka-ge(External) | |||
==Enclosure== | |||
AL-12-015Page 25 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemeAging Effect TNUREG-No TnMing Effect Aging Management 1801, Table 1No. Type Function(s) aing Program Volume 2 Item NotesM ItemTank -Anti structural Stainless Air-indoor--_Foam Tank Strutur Steele uncontrolled None None VII.J-15 3.3.1-94 C(DB-T30) i Steel (External)Tank- Collection, Drainagqe,Degasifier Structural Stainless Raw water Loss of and Treatment V11.C1-5 3.3.1-79 E-(DB-S3) integrit Steel (Internal) material Components VInspectionTank- Air withDegasifier Structural Stainless borated water None None VIIJ-16 3.3.1-99 C-(DB-S3) integrity Steel leakage(External)Tank -Degasifier Structural Stainless Air-indoor(DB-S3) ntegrity Steel uncontrolled None None VII.J-15 3.3.1-94 C(External)Tank- Collection, Drainage,Eartr Structural Stainless Raw water Loss of and TreatmentConcentrator i Steel (Internal) material Components(DB-S2) Inspection | |||
==Enclosure== | |||
AL-12-015Page 26 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended MaterialAging Management 1801, Table I NotesNo. Type Function(s) Mong Program Volume 2 Item1Management ItemTank -Air withWaste Structural Stainless borated waterCEvaporator intevrit Steel leakage None None VII.J-16 3.3.1-99 CConcentrator 6 Stel erakae,(DB-(External)Tank -Waste Air-indoorWaste Structural Stainless AridoEvaporator Strut Stail uncontrolled None None VII.J-15 3.3.1-94 CConcentrator i Steel (External)(DB-S2Tank -Structural Stainless Air-indoor A55 DWDT 1-1 integrity Steel uncontrolled None None VII.J-15 3.3.1-94 C(DB-T27) (External)Tank -Air-indoorDWDT 1-1 Structural Stainless Ai-ndo60 uncontrolled None None VII.J-15 3.3.1-94 CHold-up Tank integrity Steel (External) C(DB-T161)Tank-Miscellaneous Air-indoor65 Liquid Waste itrity Steel uncontrolled None None VII.J-15 3.3.1-94 CMonitor Tank integrity Steel (External) C(DB-T29) | |||
==Enclosure== | |||
AL-12-015Page 27 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemdAging Effect NUREG-Row Component Intended Material Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItem _Tank -Miscellaneo structural Stainless Air-indoor70 Waste Drain Stegructu Stale uncontrolled None None VII.J-15 3.3.1-94 CTank i(External)(DB-T26)Tank -Miscellaneous Ari d oWaste Structural Stainless Air-indoorA75 Waste Strity Stess uncontrolled None None VII.J-15 3.3.1-94 AEvaporator integrity Steel (External)Storage Tank(DB-T28)Tank -Radwaste structural Stainless Air-indoor A79 Vessel integrity Steel uncontrolled None None VII.J-15 3.3.1-94Skid Vessel i(External)(1 through 5)Tank -Waste84 Polishing Structural Stainless uncontrolled None None VII.J-15 3.3.1-94 ADemineralizer integrity Steel unotrolle C(DB-T125) (External) | |||
==Enclosure== | |||
AL-12-015Page 28 of 51Affected LRA SectionLRA Page No.Page 3.3-482Affected Paragraph and SentenceTable 3.3.2-269 New RowsIn response to Supplemental RAI 2.1-3 Abandoned Equipment, nine new rows are added to LRATable 3.3.2-26, "Aging Management Review Results -Service Water System," to read as follows:Table 3.3.2-26 Aging Management Review Results -Service Water SystemAging Effect NUREG-Row Component Intended Material Environment Requiring -Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 Item NItemStructural Stainless Raw water Loss of Open-Cycle CoolingOrifice ntegrity Steel (Internal) material Water VII.Cl-15 3.3.1-79 BStructural Stainless Air-indoorSuuOrifice Steel uncontrolled None None VII.J-15 3.3.1-94 A-- reintegrit Steel (External)Structural Stainless Condensation Loss of External SurfacesOrifice integrity Steel (External) material Monitoring VI-I 3.3.1-27 EStructural Stainless Raw water Loss of Open-Cycle Cooling VII. C1-15 3.3.1-79 B-- _ inte-grity Steel (Internal) material Water | |||
==Enclosure== | |||
AL-12-015Page 29 of 51Table 3.3.2-26 Aging Management Review Results -Service Water SystemAging Effect NUREG-Row Type Fntinds) Material Environment Requiring Aging Management 1801, Table INo. Type Function(s) Management Program Volume 2 Item NotesManagmentItemStructural Stainless Air-indoorStegructu Stanl uncontrolled None None VII.J-15 3.3.1-94 A-ntegrt Steel (Extemal)Structural Stainless Condensation Loss of External Surfacesinte-grity Steel (External) material Monitoring VII.Fl1-1 3.3.1-27 EStructural Stainless Raw water Loss of Open-Cycle Coolinq VII.C1-15 3.3.1-79 BValve Body integrity Steel (Internal) material WaterStructural Stainless Air-indoorSuuValve Body Steel uncontrolled None None VII.J-15 3.3.1-94 A(External)-- Valve Body Structural Stainless Condensation Loss of External Surfaces VII.FI-1 33.1-27 EBody integrity Steel (External) material Monitoring | |||
==Enclosure== | |||
AL-12-015Page 30 of 51Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable 4.1-1Page 4.1-3"Neutron Fluence" rowIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the "Neutron Fluence" row of LRA Table 4.1-1, "Time-LimitedAging Analyses," is revised as follows:Table 4.1-1 Time-Limited Aging Analyses54.21 (c)(1) LRAResults of TLAA Evaluation by Category Paragraph LRAParagraph SectionReactor Vessel Neutron Fluence 4.2Neutron Fluence N1 t .-1.T,, 4.2.120Affected LRA SectionLRA Page No.Affected Paragraph and Sentence4.2.1.3Page 4.2-3New 4th paragraph; and,"Disposition"In response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, LRA Section 4.2.1.3, "Beltline Evaluation," is revised to include anew fourth paragraph and a revised "Disposition" statement, to read as follows:Neutron fluence analysis valid for 52 EFPY have been prepared for the reactorvessel beltline materials to bound the proiected value of 50.3 EFPY for 60-yearsof operation. Therefore, the neutron fluence analysis has been proiected to theend of the period of extended operation.Disposition;. Not a TLAA Neutron fluence is an assumption uei .n various neutron embrittlement TLAA&eva'uated be/ow.Disposition: 10 CFR 54.21(c)(1)(ii) The neutron fluence analysis has beenproiected to the end of the period ofextended operation. | |||
==Enclosure== | |||
AL-12-015Page 31 of 51Affected LRA Section LRA Page No. Affected Paragraph and Sentence4.2.6 Page 4.2-14 5th paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the fifth paragraph on page 4.2-14 of LRA Section 4.2.6,"lntergranular Separation (Underclad Cracking)," is revised to read as follows:As proevidend in Confiratory' Acton Letter-, Number 3 10 001, FENOC havoluntarily comm~ittedf to shutdowný the Davis Besse plant no later- than October-1L,2011, and replace the RV closure head. Therefor~e, the cUrrent head (purchasedfrom the Midland Plant andf installed during the cycle 13 refuelhing outage) is nofconsidered in the -,nderclad cracking evaluation. The replacement RV closureheadlhead fl#ange, to be installed during the October- 2011 outage, was fabricated,usi .ng SA 508 Class 3 m~aterial, whic~h is not susceptible to interg:anUlaseparatien&- The reactor vessel closure head/head flan-ge was replaced in theFall of year 2011. This replacement head was fabricated using SA-508 Class 3material, which is not susceptible to the subiect intergranular separations.Therefore, this replacement closure head/head flange is not considered in theunderclad cracking evaluation. | |||
==Enclosure== | |||
AL-12-015Page 32 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1 Page A-9 3rd paragraphIn response to RAI B.1.4-3, the third paragraph of LRA Section A.1, "SummaryDescriptions of Aging Management Programs and Activities," previously addedby FENOC letter dated August 17, 2011 (ML11231A966), is replaced in itsentirety to read as follows:Existing FE-=NQG processes reqire rv of relevant site and industoperating exeinc n proi benchmarking to ensure proqgramea.ncrme. nt -,rz .. .ident.if..d and implemented. Such nGoi.ng rviews idenpotential need fo r- agn aagement program rew .sions to ensure thei;offoctivenes thrtoughout the period of extendfed opeFratonOperating ExperienceThe intent of evaluatinq and incorporating operating experience lessons-learned,including aging-related lessons-learned, is prevention. The lessons-learned areused to improve plant operation, equipment material condition, and agingmanagement to minimize equipment degradation and prevent loss of equipmentintended functions. The FENOC Operating Experience Program processes andprocedures for the ongoing review of operating experience include the followinqattributes:* Personnel responsible for screening, evaluating and submitting (to theindustry) aging-related operating experience items are qualified for thetask.* While the programs and procedures may specify reviews of certainsources of information, such as NRC generic communications andInstitute of Nuclear Power Operations reports, they allow for any potentialsource of relevant plant-specific or industry operating experienceinformation.The processes are adequate so as to not preclude the consideration ofoperating experience related to.-aging management. The processes allowfor appropriately gathering information on structures and passivecomponents within the scope of license renewal, their materials,environments, aging effects, and aging mechanisms, and the aging | |||
==Enclosure== | |||
AL-12-015Page 33 of 51management programs credited for managing the effects of aging,including the activities under these programs (e.g., inspection methods,preventive actions or evaluation techniques).Plant-specific operating experience, including aging-related operatingexperience, is documented in condition reports and processed using theFENOC Corrective Action Program. Condition -reports for adverseconditions and related documents captured in the Corrective ActionProgram database are quality records and are auditable and retrievable.Industry operating experience, including aging-related operatingexperience, is entered into the Operating Experience Program databaseand screened for applicability to FENOC. Documents captured in theOperating Experience Program database are retrievable.Evaluations of internal and external aging-related operating experienceissues associated with structures and passive components includeconsideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging managementprogram, with feedback to the affected aging management program ownerfor consideration of the impact to aging management programeffectiveness.Aging management program owners review data, collected by programactivities, use the Corrective Action Program to document adverseconditions to ensure they will be addressed and corrected, maintainrequired records for the program, maintain the program current, andimplement revisions as needed based on program results and internal orexternal operating experience evaluations. Revision of existing ordevelopment of new aging management programs based on operatingexperience evaluations'is performed through corrective actions using theCorrective Action Program, or by action items identified in the OperatingExperience Program database.Noteworthy plant-specific aging-related operating experience is sharedwith the other FENOC sites and the industry. The Operating ExperienceProgram procedure provides guidance on sharing internal operatingexperience. | |||
==Enclosure== | |||
AL-12-015Page 34 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.32 Page A-21 Paragraphs 1, 2, 3, 7 and 12 (last)In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals AgingManagement, the 1st, 2nd, 3rd, 7th and 12th paragraphs of LRA Section A.1.32,"PWR Reactor Vessel Internals Program," previously replaced in its entirety byFENOC letter dated September 16, 2011 (ML11264A059), are revised to readas follows (affected sentences highlighted for clarity):A.1.32 PWR REACTOR VESSEL INTERNALS PROGRAMThe PWR Reactor Vessel Internals Program relies on implementation of theElectric Power Research Institute (EPRI) Report No.10169 1022863,"Materials Reliability Program: Pressurized Water Reactor Internals Inspectionand Evaluation Guidelines (MRP-227-)," and EPRI Report No. 1016609,"Materials Reliability Program: Inspection Standard for PWR Internals(MRP-228)," to manage the aging effects on the reactor vessel internal (RVI)components.This program is used to manage the effects of age-related degradationmechanisms that are applicable in general to the PWR RVI components atDavis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effectsinclude (a) various forms of cracking, including stress corrosion cracking (SCC),which also encompasses primary water stress corrosion cracking (PWSCC),irradiation-assisted stress corrosion cracking (IASCC), or cracking due tofatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracturetoughness due to either thermal aging or neutron irradiation embrittlement; and(d) loss of preload due to thermal and irradiation-enhanced stress relaxation orcreep. In addition, the program includes management of the time-limited aginganalysis (TLAA) identified in License Renewal Application (LRA) Section A.2.2.7for reduction in fracture toughness of the reactor vessel internals. This TLAA willbe managed in accordance with the implementation of the MRP-227 guidelines,ass almendod by the A MR 227 _sfoty .va.at"on,- including all activities associatedwith Davis-Besse's responses to plant-specific action items identified inSection 4.2 of the safety evaluation.The program applies the guidance in MRP-227, 0, as am.ended by thefor inspecting, evaluating, and, if applicable, dispositioningnon-conforming RVI components at Davis-Besse. The program conforms to thedefinition of a sampling-based condition monitoring program, as defined by theBranch Technical Position RSLB-1, with periodic examinations and otherinspections of highly-affected internals locations. These examinations providereasonable assurance that the effects of age-related degradation mechanisms | |||
==Enclosure== | |||
AL-12-015Page 35 of 51will be managed during the period of extended operation. The program includesexpanding periodic examinations and other inspections if the extent of thedegradation effects exceeds the expected levels.The MRP-227 guidance for selecting RVI components for inclusion in theinspection sample is based on a four-step ranking process. Through this process,the reactor internals were assigned to one of the following four groups: Primary,Expansion, Existing Programs, and No Additional Measures components.Definitions of each group are provided in GALL Chapter IX.B.The result of this four-step sample selection process is a set of Primary InternalsComponent locations for each of the three plant designs (Westinghouse,Combustion Engineering and Babcock & Wilcox) that are expected to show theleading indications of the degradation effects, with another set of ExpansionInternals Component locations that are specified to expand the sample shouldthe indications be more severe than anticipated. The degradation effects in athird set of internals locations are deemed to be adequately managed by ExistingPrograms. A fourth set of internals locations are deemed to require no additionalmeasures. As a result, the program typically identifies 5 to 15 percent of the RVIlocations as Primary Component locations for inspections, with another 7 to10 percent of the RVI locations to be inspected as Expansion Components, aswarranted by the evaluation of the inspection results. Another 5 to 15 percent ofthe internals locations are covered by Existing Programs, with the remainderrequiring no additional measures. This process thus uses appropriate componentfunctionality criteria, age-related degradation susceptibility criteria, and failureconsequence criteria to identify the components that will be inspected under theprogram in a manner that conforms to the sampling criteria for sampling-basedcondition monitoring programs in Section A.1.2.3.4 of NRC Branch PositionRLSB-1. Consequently, the sample selection process is adequate to assure thatthe intended function(s) of the PWR reactor internal components are maintainedduring the period of extended operation.No existing generic industry programs contain the specificity considered sufficientfor monitoring the aging effects addressed by the MRP-227 guidelines for B&Wplants. Therefore, no components for B&W plants were placed into the ExistingPrograms group.MRP-227 I&E guidelines require a visual (VT-3) examination of the core supportshield (CSS) vent valve retaining for every 10 year InserviceInspection Interval. In addition, Davis-Besse Technical Specification 5.5.4requires testing of the CSS vent valves every 24 months to verify by visualinspection that the valve body and valve disc exhibit no abnormal degradation,verify the valve is not stuck in an open position, and verify by manual actuationthat the valve is fully open when a force of 5400 lbs is applied vertically upward. | |||
==Enclosure== | |||
AL-12-015Page 36 of 51The technical specification inspection will continue to be performed at theprescribed frequency of 24 months. The MRP-227 required visual (VT-3)examination will also be performed at the prescribed frequency of every 10 yearInservice Inspection Interval.The program's use of visual examination methods in MRP-227 for detection ofrelevant conditions (and the absence of relevant conditions as a visualexamination acceptance criterion) is consistent with the ASME Code, Section XIrules for visual examination. However, the program's adoption of the MRP-227guidance for visual examinations goes beyond the ASME Code, Section XI visualexamination criteria because additional guidance is incorporated into MRP-227 toclarify how the particular visual examination methods will be used to detectrelevant conditions and describes in more detail how the visual techniques relateto the specific RVI components and how to detect their applicable age-relateddegradation effects.The technical basis for detecting relevant conditions using volumetric ultrasonictesting (UT) inspection techniques can be found in MRP-228, where the review ofexisting bolting UT examination technical justifications has demonstrated theindication detection capability of at least two vendors, and where vendortechnical justification is a requirement prior to any additional boltingexaminations. Specifically, the capability of program's UT volumetric methods todetect loss of integrity of PWR internals bolts, pins, and fasteners, such asbaffle-former bolting in B&W and Westinghouse units, has been welldemonstrated by operating experience. In addition, the program's adoption of theMRP-227 guidance and process incorporates the UT criteria in MRP-228, whichcalls for the technical justifications that are needed for volumetric examinationmethod demonstrations, required by the ASME Code, Section V.The program also includes future industry operating experience as incorporatedin periodic revisions to MRP-227. The. program thus provides reasonableassurance for the long-term integrity and safe operation of reactor internals in allcommercial operating U.S. PWR nuclear power plants.Age-related degradation in the reactor internals is managed through anintegrated program. Specific features of the integrated program are listed in thefollowing ten program elements. Degradation due to changes in materialproperties (e.g., loss of fracture toughness) was considered in the determinationof inspection recommendations and is managed by the requirement to useappropriately degraded properties in the evaluation of identified defects. Theintegrated program is implemented by the applicant through an inspection plan.The Davis-Besse PWR Reactor Vessel Internals Program will address allplant-specific action items applicable to Davis-Besse that are established in | |||
==Enclosure== | |||
AL-12-015Page 37 of 51Section 4.2 of the safety evaluation for MRP-227. In addition, a plant-specificinspection plan for ensuring the implementation of MRP-227 program guidelines,a. amended by the safoty 9a-uatiop for. and Davis-Besse's responsesto the plant-specific action items, as identified in Section 4.2 of the safetyevaluation for MR 22.", will be submitted for NRC review and approval.Affected LRA Section LRA Page No.Affected Para-graph and SentenceA.2.2.1Pages A-30and A-31Ist paragraph; andNew last paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 1st paragraph of LRA Section A.2.2.1, "Neutron Fluence," isdeleted, and a new last paragraph is added to the Section on LRA page A-31,as follows:1st Para-graphNeutroen flence is not a T-14, it is a timo limited assumptin used in theevaluation of neutron embrittlement TLAAs.New Last ParagraphA neutron fluence analysis valid for 52 EFPY has been prepared for the reactorvessel beltline materials to bound the proiected value of 50.3 EFPY for60 years of operation. Therefore, the neutron fluence analysis has beenproiected to the end of the period of extended operation in accordance with10 CFR 54.21(c)(1)(ii). | |||
==Enclosure== | |||
AL-12-015Page 38 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.2.2.6 Page A-34 3rd paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 3rd paragraph of LRA Section A.2.2.6, "Neutron Fluence," isrevised to read as follows:[Pro posed for this sc.tibn, pendnIg of Action Ltte:C41Ain 2o 10 001 conmmitments related to theg ,olcroto h ai ashead in 2011.] The reactor vessel closure head/head flange wasreplaced in the fall of year 2011. This replacement head was fabricated usingSA-508 Class 3 material, which is not susceptible to the subject intergranularseparations. Therefore, this replacement closure head/head flange is notconsidered in the underclad cracking evaluation. | |||
==Enclosure== | |||
AL-12-015Page 39 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.29 Page B-119 2nd ParagraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 2nd paragraph of the "Operating Experience" subsection ofLRA Section B.2.29, "Nickel-Alloy Reactor Vessel Closure Head NozzlesProgram," on LRA page B-119, is revised as follows:In March 2010, ultrasonic examinations of the control rod drive mechanismnozzles constructed of Alloy 600 material identified flaws on multiple nozzles.Active leakage was identified on one nozzle. The direct cause was PrimaryWater Stress Corrosion Cracking. The reactor vessel closure head had been inoperation approximately six years. An inside diameter temper bead half-nozzleweld repair was utilized. Post-repair inspections were completed with acceptableresults. As pro vided in Action Numbe ^h r: -3 10- 060 4, MaFk A.Satorius (NR1) to a-ry S. Allen dated 66 23 22010, hhavoluntaril committed to shutdown the Davis B esse plant no later- than October- 1,20114, and replace the reactor pressuq~re_ 14esse90 head- With one man1ufacturFed usingmaterials rosistant to PWSCC. The reactor vessel closure head was replacedwith a new head in the fall of year 2011. The CRDM nozzles for the new headwere fabricated using Alloy 690 material that is less susceptible to PWSCC. | |||
==Enclosure== | |||
AL-12-015Page 40 of 51Affected LRA Section LRA Page No. Affected Para-graph and SentenceB.2.32 Pages B-129 Various paragraphsthru B-133In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals AgingManagement, various paragraphs of LRA Section B.2.32, "PWR Reactor VesselInternals Program," previously replaced in its entirety by FENOC letter datedSeptember 16, 2011 (ML1 1264A059), are revised to read as follows (affectedsentences highlighted for clarity):B.2.32 PWR REACTOR VESSEL INTERNALS PROGRAMProgram DescriptionThe PWR Reactor Vessel Internals Program relies on implementation of theElectric Power Research Institute (EPRI) Report No.10659 1022863,"Materials Reliability Program: Pressurized Water Reactor Internals Inspectionand Evaluation Guidelines (MRP-227-.A)," and EPRI Report No. 1016609,"Materials Reliability Program: Inspection Standard for PWR Internals(MRP-228)," to manage the aging effects on the reactor vessel internal (RVI)components.This program is used to manage the effects of age-related degradationmechanisms that are applicable in general to the PWR RVI components atDavis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effectsinclude (a) various forms of cracking, including stress corrosion cracking (SCC),which also encompasses primary water stress corrosion cracking (PWSCC),irradiation-assisted stress corrosion cracking (IASCC), or cracking due tofatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracturetoughness due to either thermal aging or neutron irradiation embrittlement; and(d) loss of preload due to thermal and irradiation-enhanced stress relaxation orcreep. In addition, the program includes management of the time-limited aginganalysis (TLAA) identified in License Renewal Application (LRA) Section 4.2.7 forreduction in fracture toughness of the reactor vessel internals. This TLAA will bemanaged in accordance with the implementation of the MRP-227 guidelinesr-asa.meonedad by tho MRP 227- safty including all activities associatedwith Davis-Besse's responses to plant-specific action items identified inSection 4.2 of the safety evaluation.The program applies the guidance in MRP-227, Rev. 0, as am.en ed by th'for inspecting, evaluating, and, if applicable, dispositioningnon-conforming RVI components at Davis-Besse. The program conforms to thedefinition of a sampling-based condition monitoring program, as defined by the | |||
==Enclosure== | |||
AL-12-015Page 41 of 51Branch Technical Position RSLB-1, with periodic examinations and otherinspections of highly-affected internals locations. These examinations providereasonable assurance that the effects of age-related degradation mechanismswill be managed during the period of extended operation. The program includesexpanding periodic examinations and other inspections if the extent of thedegradation effects exceeds the expected levels.The MRP-227 guidance for selecting RVI components for inclusion in theinspection sample is based on a four-step ranking process. Through this process,the reactor internals were assigned to one of the following four groups: Primary,Expansion, Existing Programs, and No Additional Measures components.Definitions of each group are provided in GALL Chapter IX.B.The result of this four-step sample selection process is a set of Primary InternalsComponent locations for each of the three plant designs (Westinghouse,Combustion Engineering and Babcock & Wilcox) that are expected to show theleading indications of the degradation effects, with another set of ExpansionInternals Component locations that are specified to expand the sample shouldthe indications be more severe than anticipated. The degradation effects in athird set of internals locations are deemed to be adequately managed by ExistingPrograms. A fourth set of internals locations are deemed to require no additionalmeasures. As a result, the program typically identifies 5 to 15 percent of the RVIlocations as Primary Component locations for inspections, with another 7 to10 percent of the RVI locations to be inspected as Expansion Components, aswarranted by the evaluation of the inspection results. Another 5 to 15 percent ofthe internals locations are covered by Existing Programs, with the remainderrequiring no additional measures. This process thus uses appropriate componentfunctionality criteria, age-related degradation susceptibility criteria, and failureconsequence criteria to identify the components that will be inspected under theprogram in a manner that conforms to the sampling criteria for sampling-basedcondition monitoring programs in Section A.1.2.3.4 of NRC Branch PositionRLSB-1. Consequently, the sample selection process is adequate to assure thatthe intended function(s) of the PWR reactor internal components are maintainedduring the period of extended operation.No existing generic industry programs contain the specificity considered sufficientfor monitoring the aging effects addressed by the MRP-227 guidelines for B&Wplants. Therefore, no components for B&W plants were placed into the ExistingPrograms group.The program's use of visual examination methods in MRP-227 for detection ofrelevant conditions (and the absence of relevant conditions as a visualexamination acceptance criterion) is consistent with the ASME Code, Section Xlrules for visual examination. However, the program's adoption of the MRP-227 | |||
==Enclosure== | |||
AL-12-015Page 42 of 51guidance for visual examinations goes beyond the ASME Code, Section Xl visualexamination criteria because additional guidance is incorporated into MRP-227 toclarify how the particular visual examination methods will be used to detectrelevant conditions and describes in more detail how the visual techniques relateto the specific RVI components and how to detect their applicable age-relateddegradation effects.The technical basis for detecting relevant conditions using volumetric ultrasonictesting (UT) inspection techniques can be found in MRP-228, where the review ofexisting bolting UT examination technical justifications has demonstrated theindication detection capability of at least two vendors, and where vendortechnical justification is a requirement prior to any additional boltingexaminations. Specifically, the capability of program's UT volumetric methods todetect loss of integrity of PWR internals bolts, pins, and fasteners, such asbaffle-former bolting in B&W and Westinghouse units, has been welldemonstrated by operating experience. In addition, the program's adoption of theMRP-227 guidance and process incorporates the UT criteria in MRP-228, whichcalls for the technical justifications that are needed for volumetric examinationmethod demonstrations, required by the ASME Code, Section V.The program also includes future industry operating experience as incorporatedin periodic revisions to MRP-227. The program thus provides reasonableassurance for the long-term integrity and safe operation of reactor internals in allcommercial operating U.S. PWR nuclear power plants.Age-related degradation in the reactor internals is managed through anintegrated program. Specific features of the integrated program are listed in thefollowing ten program elements. Degradation due to changes in materialproperties (e.g., loss of fracture toughness) was considered in the determinationof inspection recommendations and is managed by the requirement to useappropriately degraded properties in the evaluation of identified defects. Theintegrated program is implemented by the applicant through an inspection plan.The Davis-Besse PWR Reactor Vessel Internals Program will address allplant-specific action items applicable to Davis-Besse that are established inSection 4.2 of the safety evaluation for MRP-227. In addition, a plant-specificinspection plan for ensuring the implementation of MRP-227 program guidelinesT,as amended by tho safoy oval.ati,,n Afor AdRip 227-, and Davis-Besse's responsesto the plant-specific action items, as identified in Section 4.2 of the safetyevaluationI ,Ar- MR2.2-J, will be submitted for NRC review and approval. | |||
==Enclosure== | |||
AL-12-015Page 43 of 51NUREG-1801 ConsistencyThe PWR Reactor Vessel Internals Program is a new Davis-Besse program thatwill be consistent with the 10 elements of an effective aging managementprogram as described in NUREG-1801, Rev. 2, Section XI.M16A, "PWR VesselInternals." The results of an evaluation for each element are provided below.Exceptions to NUREG-1801None.EnhancementsNone.Aging Management Program ElementsThe results of an evaluation of each program element are provided below.ScopeThe scope of the program includes all RVI components at Davis-Besse, whichis built to a B&W NSSS design. The scope of the program applies themethodology and guidance in the most recently NRC-endorsed version ofMRP-227, which provides augmented inspection and flaw evaluationmethodology for assuring the functional integrity of safety-related internals incommercial operating U.S. PWR nuclear power plants designed by B&W, CE,and Westinghouse. The scope of components considered for inspectionunder MRP-227 guidance includes core support structures (typically denotedas Examination Category B-N-3 by the ASME Code, Section XI), those RVIcomponents that serve an intended license renewal safety function pursuantto criteria in 10 CFR 54.4(a)(1), and other RVI components whose failurecould prevent satisfactory accomplishment of any of the functions identified in10 CFR 54.4(a)(1)(i), (ii), or (iii). The scope of the program does not includeconsumable items, such as fuel assemblies, reactivity control assemblies,and nuclear instrumentation, because these components are not typicallywithin the scope of the components that are required to be subject to an agingmanagement review (AMR), as defined by the criteria set in10 CFR 54.21(a)(1). The scope of the program also does not include weldedattachments to the internal surface of the reactor vessel because thesecomponents are considered to be ASME Code Class 1 appurtenances to thereactor vessel and are adequately managed in accordance with an applicant'saging management program (AMP) that corresponds to GALL AMP XI.M1,"ASME Code, Section XI Inservice Inspection, Subsections IWB, IWC,and IWD." | |||
==Enclosure== | |||
AL-12-015Page 44 of 51In addition, the scope of the program includes management of thetime-limited aging analysis (TLAA) identified in LRA Section 4.2.7 forreduction in fracture toughness of the reactor vessel internals. This TLAA willbe managed in accordance with the implementation of the MRP-227guidelines, as amendod by thoeMRP 227 Safoty ovaluation, including allactivities associated with Davis-Besse's responses to plant-specific actionitems identified in the Section 4.2 of the safety evaluation.The scope of the program includes the response bases to applicable licenserenewal applicant action items (LRAAIs) on the MRP-227 methodology, andany additional programs, actions, or activities that are discussed in theseLRAAI responses and credited for aging management of the applicant's RVIcomponents. The LRAAIs are identified in the staffs safety evaluation onMRP-227 and include applicable action items on meeting those assumptionsthat formed the basis of the MRP's augmented inspection and flaw evaluationmethodology (as discussed in Section 2.4 of MRP-227), and NSSSvendor-specific or plant-specific LRAAIs as well. Davis-Besse's responses tothe plant-specific action items, as identified in Section 4.2 of the safetyevaluation for MRP-227, will be submitted for NRC review and approval.The guidance in Section 2.4 of MRP-227 specifies applicability limitations tobase-loaded plants and the fuel loading management assumptions uponwhich the functionality analyses were based. General assumptions used inthe analysis include:1) 30 years of operation with high leakage core loading patterns (freshfuel assemblies loaded in peripheral locations) followed byimplementation of a low-leakage fuel management strategy for theremaining 30 years of operation;2) base load operation, i.e., typically operates at fixed power levels anddoes not usually vary power on a calendar or load demand schedule;and3) no design changes beyond those identified in general industryguidance or recommended by the original vendors.Davis-Besse had approximately 13 years of operation with fresh fuelassemblies at peripheral locations. Cycle 15 has implemented a new failureresistant fuel design in high vulnerability locations. The core design forDavis-Besse is within the assumption of MRP-227. Davis-Besse is a baseload plant and has incorporated no design changes beyond those identified ingeneral industry guidance or recommended by the original vendors. | |||
==Enclosure== | |||
AL-12-015Page 45 of 51* Preventive ActionsThe guidance in MRP-227 relies on PWR water chemistry control to preventor mitigate aging effects that can be induced by corrosive aging mechanisms(e.g., loss of material induced by general, pitting corrosion, crevice corrosion,or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]).Reactor coolant water chemistry is monitored and maintained in accordancewith the PWR Water Chemistry Program. The PWR Water ChemistryProgram is an existing Davis-Besse program that is consistent with the10 elements of an effective aging management program as described inNUREG-1801, Section XI.M2, "Water Chemistry."* Parameters Monitored or InspectedThe program manages the following age-related degradation effects andmechanisms that are applicable in general to the RVI components at thefacility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclicalloading; (b) loss of material induced by wear; (C) loss of fracture toughnessinduced by either thermal aging or neutron irradiation embrittlement;(d) changes in dimension due to void swelling and irradiation growth,distortion, or deflection; and (e) loss of preload caused by thermal andirradiation-enhanced stress relaxation or creep. For the management ofcracking, the program monitors for evidence of surface breaking lineardiscontinuities if a visual inspection technique is used as the non-destructionexamination (NDE) method, or for relevant flaw presentation signals if avolumetric UT method is used as the NDE method. For the management ofloss of material, the program monitors for gross or abnormal surfaceconditions that may be indicative of loss of material occurring in thecomponents. For the management of loss of preload, the program monitorsfor gross surface conditions that may be indicative of loosening in applicablebolted, fastened, keyed, or pinned connections. The program does notdirectly monitor for loss of fracture toughness that is induced by thermal agingor neutron irradiation embrittlement, or by void swelling and irradiation growth;instead, the impact of loss of fracture toughness on component integrity isindirectly managed by using visual or volumetric examination techniques tomonitor for cracking in the components and by applying applicable reducedfracture toughness properties in the flaw evaluations if cracking is detected inthe components and is extensive enough to warrant a supplemental flawgrowth or flaw tolerance evaluation under the MRP-227 guidance or ASMECode, Section Xl requirements. The program uses physical measurements tomonitor for any dimensional changes due to void swelling, irradiation growth,distortion, or deflection.Specifically, the program implements the parameters monitored/inspectedcriteria for B&W designed Primary Components in Table 4-1 of MRP-227. | |||
==Enclosure== | |||
AL-12-015Page 46 of 51Additionally, the program implements the parameters monitored/inspectedcriteria for B&W designed Expansion Components in Table 4-4 of MRP-227.No existing generic industry programs contain the specificity consideredsufficient for monitoring the aging effects addressed by the MRP-227guidelines for B&W plants. Therefore, no components for B&W plants wereplaced into the Existing Programs group. No inspections, except for thosespecified in ASME Code, Section Xl, are required for components that areidentified as requiring "No Additional Measures," in accordance with theanalyses reported in MRP-227. As part of the Davis-Besse InserviceInspection Program, a visual VT-3 examination of the reactor vesselremovable core support structure is conducted once per Inservice Inspectioninterval in accordance with ASME Section Xl, Table IWB-2500-1, ExaminationCategory B-N-3.MRP-227 I&E guidelines require a visual (VT-3) examination of the coresupport shield (CSS) vent valve retaining rings and disc ahaft for every10year Inservice Inspection Interval. In addition, Davis-Besse TechnicalSpecification 5.5.4 requires testing of the CSS vent valves every 24 months toverify by visual inspection that the valve body and valve disc exhibit noabnormal degradation, verify the valve is not stuck in an open position, andverify by manual actuation that the valve is fully open when a force of<400 lbs is applied vertically upward. The technical specification inspectionwill continue to be performed at the prescribed frequency of 24 months. TheMRP-227 required visual (VT-3) examination will also be performed at theprescribed frequency of every 10 year Inservice Inspection Interval.Detection of Aging EffectsThe detection of aging effects is covered in two places: (a) the guidance inSection 4 of MRP-227 provides an introductory discussion and justification ofthe examination methods selected for detecting the aging effects of interest;and (b) standards for examination methods, procedures, and personnel areprovided in a companion document, MRP-228. In all cases, well-establishedmethods were selected. These methods include volumetric UT examinationmethods for detecting flaws in bolting, physical measurements for detectingchanges in dimension, and various visual (VT-3, VT-1, and EVT-1)examinations for detecting effects ranging from general conditions todetection and sizing of surface-breaking discontinuities. Surface examinationsmay also be used as an alternative to visual examinations for detection andsizing of surface-breaking discontinuities.Cracking caused by SCC, IASCC, and fatigue is monitored/inspected byeither VT-1 or EVT-1 examination (for internals other than bolting) or byvolumetric UT examination (bolting). The VT-3 visual methods may be applied | |||
==Enclosure== | |||
AL-12-015Page 47 of 51for the detection of cracking only when the flaw tolerance of the component oraffected assembly, as evaluated for reduced fracture toughness properties, isknown and has been shown to be tolerant of easily detected large flaws, evenunder reduced fracture toughness conditions. In addition, VT-3 examinationsare used to monitor/inspect for loss of material induced by wear and forgeneral aging conditions, such as gross distortion caused by void swellingand irradiation growth or by gross effects of loss of preload caused by thermaland irradiation-enhanced stress relaxation and creep.In addition, the program adopts the recommended guidance in MRP-227 fordefining the Expansion criteria that need to be applied to inspections ofPrimary Components and Existing Requirement Components and forexpanding the examinations to include additional Expansion Components. Asa result, inspections performed on the RVI components are performedconsistent with the inspection frequency and sampling bases for PrimaryComponents and Expansion Components in MRP-227, which have beendemonstrated to be in conformance with the inspection criteria, samplingbasis criteria, and sample Expansion criteria in Section A.1.2.3.4 of NRCBranch Position RLSB-1.Specifically, the program implements the parameters monitored/inspectedcriteria and bases for inspecting the relevant parameter conditions for B&Wdesigned Primary Components in Table 4-1 of MRP-227 and for B&Wdesigned Expansion Components in Table 4-4 of MRP-227.As provided in Soctien 4.4.3 of the MRP 227 sapey evalan, theasrme'disrbutse t sho l efotaging bolte (aiso known as the flow destebutsr boes) inRA 144 designed plants Wares dded to the relmain inspegtien d ategnsio Thesafety evaluatien pro-vides-Q th-at the ex-aminationa method- shaoll bhe volume~tricexaminatid n (UT), the examnnatien coverage for these aemponents shaolponfo to the acrieria as desrogbed in Section 3.3.41 of the saof evalRat-n,and the Fe examination frequency shall be on a 40 year- int-ernal sAimilar to1olther 'PrhMary" inspection caegory componsents. For BR&W designed plants,,Pno other a;;dditinal_ components were- added- to the "Priar-y" insectioncategor-y and no add-itional ~omponents Iere a~dded to the- "Expansion"insection categeor/In addition, in some cases (as defined in MVRP-227), physical measurementsare used as supplemental techniques to manage for the gross effects of wear,loss of preload due to stress relaxation, or for changes in dimension due tovoid swelling, deflection or distortion. The physical measurements methodsapplied in accordance with this program includes Section 4.3.1 of MRP-227that describes the physical measurements needed for the B&W internals coreclamping items. In addition, Table 4-1 provides the required examination | |||
==Enclosure== | |||
AL-12-015Page 48 of 51method and examination coverage and Table 5-1 provides the acceptancecriteria for the physical measurements.Monitoring and TrendingThe program requires that all inspections shall be documented for futurereview; defects shall be documented in accordance with the Davis-Bessecorrective action program.In addition, the program requires that a summary report of all inspections andmonitoring, items requiring evaluation, and new repairs shall be submitted tothe MRP Program Manager within 120 days of the completion of an outageduring which PWR internals within the scope of MRP-227 are examined.Section 6 of MRP-227 will not be used by FENOC for evaluating examinationresults that do not meet the acceptance criteria identified in Section 5 ofMRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as theframework to develop those generic and plant-specific evaluations triggeredby findings in the RVI examinations. As provided in the safety evaluation forMRP-227, Rev-0-, the NRC staff is currently reviewing WCAP-17096-NP,Revision 2." Acceptance CriteriaSection 5 of MRP-227 provides specific examination acceptance criteria forthe Primary and Expansion Component examinations. For componentsaddressed by examinations referenced to ASME Code, Section Xl, theIWB-3500 acceptance criteria apply. For other components covered byExisting Programs, the examination acceptance criteria are described withinthe Existing Program reference document.The guidance in MRP-227 contains three types of examinationacceptance criteria:For visual examination (and surface examination as an alternative tovisual examination), the examination acceptance criterion is theabsence of any of the specific, descriptive relevant conditions; inaddition, there are requirements to record and disposition surfacebreaking indications that are detected and sized for length byVT-1/EVT-1 examinations;For volumetric examination, the examination acceptance criterion isthe capability for reliable detection of indications in bolting, asdemonstrated in the examination Technical Justification; in addition,there are requirements for system-level assessment of bolted or | |||
==Enclosure== | |||
AL-12-015Page 49 of 51pinned assemblies with unacceptable volumetric (UT) examinationindications that exceed specified limits; and,For physical measurements, the examination acceptance criterion forthe acceptable tolerance in the measured differential height from thetop of the plenum rib pads to the vessel seating surface in B&W plantsare given in Table 5-1 of MRP-227.Section 6 of MRP-227 will not be used by FENOC for evaluating examinationresults that do not meet the acceptance criteria identified in Section 5 ofMRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as theframework to develop those generic and plant-specific evaluations triggeredby findings in the RVI examinations. As provided in the safety evaluation forMRP-227, Rev- OL, the NRC staff is currently reviewing WCAP-17096-NP,Revision 2." 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 programs and activities that arecredited with aging management during the period of extended operation andis discussed in Section B .1.3.* Operating ExperienceRelatively few incidents of PWR internals aging degradation have beenreported in operating U.S. commercial PWR plants. However, a considerableamount of PWR internals aging degradation has been observed in EuropeanPWRs, with emphasis on cracking of baffle-former bolting. For this reason,the U.S. PWR owners and operators began a program a decade ago toinspect the baffle-former bolting in order to determine whether similarproblems might be expected in U.S. plants. A benefit of this decision was theexperience gained with the UT examination techniques used in theinspections. In addition, the industry began substantial laboratory testingprojects in order to gather the materials data necessary to support futureinspections and evaluations. Another item with existing or suspected material | |||
==Enclosure== | |||
AL-12-015Page 50 of 51degradation concerns that has been identified for PWR components iscracking in some high-strength bolting. This condition has been correctedprimarily through bolt replacement with less susceptible material andimproved control of pre-load.Stress corrosion cracking (SCC) has occurred in Alloy A-286 internals boltingin B&W units, this included Davis-Besse. The Alloy A-286 bolt failures in B&WPWR internals were subjected to a comprehensive failure analysis that isdocumented in BAW-1843PA, "The B&W Owners Group Evaluation ofInternal Bolting Concerns in 177FA Plants," dated January 1986.BAW-1843PA was reviewed and approved by the NRC. This failure analysisaddressed probable cause of the cracking, assessment of likelihood andconsequences of joint failure, and replacement bolt design. Therecommended replacement bolts were Alloy X-750 HTH bolts that are lesssusceptible to SCC and have overall excellent material properties.Davis-Besse has replaced the majority of the Alloy A-286 bolts for the reactorvessel internals (upper core barrel, lower core barrel, lower thermal shieldand surveillance specimen holder tubes) with Alloy X-750 HTH bolts. Tosatisfy a needed action under NEI 03-08 protocol, Davis-Besse performed UTexaminations of 100% of all upper core barrel bolts during the cycle 16refueling outage. This inspection did not identify any unacceptableindications.As part of the Inservice Inspection Program, a visual (VT-3) examination ofthe reactor vessel removable core support structure is conducted once perInservice Inspection interval in accordance with ASME Section Xl, TableIWB-2500-1, Examination Category B-N-3. These inspections have notidentified any unacceptable indications.FENOC participates in the industry programs for investigating and managingaging effects on reactor vessel internals. Through its participation in EPRIMRP activities, FENOC will continue to benefit from the reporting of reactorvessel internals inspection information, and will share its own internalsinspection results with the industry, as appropriate.ConclusionThe PWR Reactor Vessel Internals Program provides reasonable assurance thatcracking, including stress corrosion cracking (SCC), which also encompassesprimary water stress corrosion cracking (PWSCC), irradiation-assisted stresscorrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; loss ofmaterial induced by wear; loss of fracture toughness due to either thermal agingor neutron irradiation embrittlement; and loss of preload due to thermal and | |||
==Enclosure== | |||
AL-12-015Page 51 of 51irradiation-enhanced stress relaxation or creep of the subject reactor vesselinternals components will be adequately managed so that intended functions ofcomponents within the scope of license renewal are maintained consistent withthe current licensing basis for the period of extended operation. | |||
==Enclosure== | |||
BDavis-Besse Nuclear Power Station, Unit No. I (DBNPS)Letter L-12-015New and Revised DBNPS License Renewal Application Boundarv Drawinqs14 pages followThe following License Renewal Application Boundary Drawingsare new and are enclosed:LR Drawing LR-MO10D Sheet 2LR Drawing LR-M036C Sheet 2LR Drawing LR-M039B Sheet 2Revision 0Revision 0Revision 0The following License Renewal Application Boundary Drawingsare revised and are enclosed:LR Drawing LR-MO10D Sheet ILR Drawing LR-MO12E,LR Drawing LR-M020BLR Drawing LR-M036C Sheet ILR Drawing LR-M037CLR Drawing LR-M037DLR Drawing LR-M037ELR Drawing LR-M037FLR Drawing LR-M039ALR Drawing LR-M039B Sheet 1LR Drawing LR-M042BRevision 3Revision 2Revision 3Revision 3Revision 2Revision 4Revision 3Revision 2Revision 4Revision 3Revision 3 The 14 Drawingsspecifically. referenced inEnclosure B have beenprocessed into ADAMSThese drawings can beaccessed by the NRC staffwithin the ADAMS packageor by performing a searchon the Document/ReportNumberr' D01=D14 | |||
}} | }} | ||
Revision as of 05:42, 2 April 2018
| ML12094A383 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/09/2012 |
| From: | Allen B S FirstEnergy Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-12-015, TAC ME4640 | |
| Download: ML12094A383 (82) | |
Text
FENOCFirstEnergy Nuclear Operating Company5501 North State Route 2Oak Harbor. Ohio 43449Barry S. AllenVice President -Nuclear419-321-7676Fax: 419-321-7582March 9, 2012L-12-01510 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), LicenseRenewal Application Amendment No. 24, and Revised License Renewal ApplicationBoundary DrawingsBy 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 letter dated December 27, 2011(ML 1333A396), the Nuclear Regulatory Commission (NRC) requested additionalinformation to complete its review of the License Renewal Application (LRA).The NRC letter contained four requests for additional information (RAIs). The FENOCresponse to one of the four RAI responses, RAI 3.1.2.2.16-3, was submitted to the NRCby letter dated January 13, 2011 (ML12018A338).Attachment 1 provides the FENOC responses to two of the four RAIs (B.1.4-2 andB.1.4-3) in the NRC letter. The submittal date for these responses was extendedfollowing discussion with Mr. Samuel Cuadrado de Jesus, NRC Project Manager,because additional time was needed for development of the responses due tocoordination with multiple site and Fleet departments. Attachment 1 also providessupplemental information on the topics listed below. The NRC request is shown in boldtext followed by the FENOC response:* RAI 2.1-3 regarding abandoned equipment;* RAI 3.1.2.2-2 regarding Reactor Vessel Internals aging management; and* LRA Section 4.2, "Reactor Vessel Neutron Embrittlement." az,It-f1I "H,1-1- ý-
Davis-Besse Nuclear Power Station, Unit No. 1L-12-015Page 2Enclosure A provides Amendment No. 24 to the DBNPS LRA. Enclosure B providesnew and revised License Renewal Boundary Drawings.The FENOC response to the fourth of four RAIs (B.2.39-13) in the NRC letter is plannedto be provided to the NRC by March 30, 2012, due to the need to evaluate andincorporate information from the Davis-Besse Shield Building concrete cracking RootCause Analysis Report that was recently completed and submitted to the NRC byFENOC letter dated February 27, 2012 (ML1 20600056).Attachment 2 identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)in this document. Any other actions discussed in the submittal represent intended orplanned actions by FENOC; they are described only as information and are notRegulatory Commitments. Please notify Mr. Clifford I. Custer, Project Manager- FleetLicense Renewal, at (724) 682-7139 of any questions regarding this document orassociated Regulatory Commitments.I declare under penalty of perjury that the foregoing is true and correct. Executed onMarch 1¶ , 2012.Sincerely,Barry S. AllenAttachment:1. Reply to Requests for Additional Information for the Review of the Davis-BesseNuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,Sections B.1.4, 2.1, 3.1.2 and 4.22. Regulatory Commitment ListEnclosures:A. Amendment No. 24 to the DBNPS License Renewal ApplicationB. New and Revised DBNPS License Renewal Application Boundary Drawingscc: NRC DLR Project ManagerNRC Region III Administrator Davis-Besse Nuclear Power Station, Unit No. 1L-12-015Page 3cc: w/o Attachment or EnclosuresNRC DLR DirectorNRR DORL Project ManagerNRC Resident InspectorUtility Radiological Safety Board
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1L-12-015Reply to Requests for Additional Information for the Review of the.Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application,Sections B.1.4, 2.1, 3.1.2 and 4.2Page 1 of 24Section B.1.4Question RAI B.1.4-2Background:In request for additional information (RAI) B.1.4-1, issued on May 19, 2011, thestaff asked the applicant to describe the programmatic activities that will be usedto continually identify aging issues, evaluate them, and as necessary, enhance theaging management programs (AMPs) or develop new AMPs for license renewal. Inits response dated June 24, 2011, the applicant stated that it currently has aprocedurally controlled operating experience review process, as required byNUREG-0737, "Clarification of TMI Action Plan Requirements," Item I.C.5,"Procedures for Feedback of Operating Experience to Plant Staff." The applicantstated that this process provides for the systematic identification and transfer oflessons learned from site and industry experience into-fleet and station processesto prevent events and enhance the safety'and reliability of its operations.Issue:The applicant's response provided a general description of how it considersoperating experience on an ongoing basis; however, it does not directly addressseveral areas in RAI B.1.4-1 on which the staff requested information. Further, theapplicant's response did not provide specific information on how the operatingexperience review activities address issues specific to aging. The staff identifiedthe following issues with the applicant's response:(a) The applicant did not describe the sources of plant-specific operatingexperience information that it monitors on an ongoing basis. Additionaldetails are needed to determine whether the applicant will consider anadequate scope of information from which to identify potential operatingexperience related to aging.(b) It is not clear whether the applicant only reviews certain sources foroperating experience information. Additional information is needed todetermine whether the applicant's processes would preclude theconsideration of relevant operating experience Information, because it isnot from a prescribed source.
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1L-12-015Page 2 of 24(c) The staff requested that the applicant indicate which guidance documentsrequire monitoring. Theapplicant did not indicate whether it considersguidance documents to be a source of operating experience information.(d) The applicant did not describe its criteria for identifying and categorizingoperating experience items as related to aging.(e) The staff requested that the applicant describe training provided to plantpersonnel. The applicant did not describe training on aging issues, nor didit indicate whether the training will be provided for those plant personnelresponsible for screening, assigning, evaluating, and submitting operatingexperience items.(f) The applicant did not describe how evaluations of operating experiencerelated to aging consider the potentially affected plant* systems, structures, and components,4 materials,* environments,* aging effects,* aging mechanisms, and" AMPs.(g) The applicant did not describe how it will consider as operating experiencethe results of the inspections, tests, analyses, etc., conducted throughimplementation of the AMPs.(h) The applicant did not describe the records of operating experienceevaluations or how it retains those records.(i) The applicant stated that operating experience evaluations are prioritizedwith due dates procedurally specified based on the potential significanceof the issue; however, the applicant did not provide details on theevaluation schedules or how it determines the relative significance ofissues. It is therefore unclear whether the operating experienceevaluations will be completed in a timely manner or whether they will be-appropriately prioritized.(j) The applicant stated that it enters operating experience that potentiallyrepresents~a condition adverse to quality into the corrective actionprogram; however, the applicant did not explain how a "condition adverseto quality" includes aging. Additional information, is needed to determine
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1L-12-015Page 3 of 24whether the corrective action program has a threshold appropriate tocapture items related to aging.(k) The applicant did not describe criteria for considering when AMPs shouldbe modified or new AMPs developed due to operating experience. It alsodid not describe how it implements these kinds of changes or how itensures the changes are implemented in a timely manner.(I) The applicant stated that it shares lessons learned with other utilities topromote industry-wide safety and reliability; however, the applicant did notprovide criteria for reporting its plant-specific operating experience onage-related degradation to the industry.Request:Provide a response to each item below.(a) Describe the sources of plant-specific operating experience that aremonitored on an ongoing basis to identify potential aging issues.(b) Indicate whether plant-specific and industry operating experience is onlyconsidered from a prescribed list of sources. If only prescribed sources areconsidered, provide ajustification as to why it is unnecessary to considerother sources.(c) Indicate whether guidance documents are considered as a source ofoperating experience information. If they are considered as a potentialsource, provide a plan for considering the content of guidancedocuments, such as the GALL Report,, as operating experience applicableto aging management.(d) Describe how operating experience issues will be identified andcategorized as related to aging.(e) Describe the training requirements on aging issues for those plantpersonnel responsible for screening, evaluating, and submitting operatingexperience items.(f) Describe how evaluations of operating experience issues related to agingwill consider the following:* systems, structures, or componentsa materials* environments0 aging effect
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1L-12-015Page 4 of 24* aging mechanisms* AMPs(g) Describe how the results of the AMP inspections, tests, analyses, etc., willbe considered as operating experience.(h) Describe the operating experience evaluation records with respect to whatis considered for aging. Indicate whether-these records are maintained inauditable and retrievable form.(i) Provide details on the operating experience evaluation schedules andjustify why they provide for timely evaluations. Also, describe how therelative significance of operating experience items is determined so thatthe reviews can be prioritized appropriately.(j) Justify why the corrective action program has an appropriate threshold forcapturing issues concerning aging.(k) Describe the criteria for considering when AMPs should be modified ornew AMPs developed due to operating experience. Also, describe theprocess for implementing changes to the AMPs or for implementing newAMPs; describe how these changes are implemented in a timely manner.(I) Provide criteria for reporting plant-specific operating experience onage-related degradation to the industry.If enhancements are necessary, provide an implementation schedule forincorporating them into the existing programmatic operating experiencereview activities.RESPONSE RAI B.1.4-2(a) Describe the sources of plant-specific operating experience that are monitoredon an ongoing basis to identify potential aging issues.The sources of plant-specific operating experience to identify potential aging issuesinclude the following:-- FENOC Corrective Action Program -adverse conditions are documented inthe Corrective Action Program, including, when appropriate, the cause andactions necessary to correct and/or prevent recurrence. Adverse conditionsas defined in the Corrective Action Program include any event, defect,characteristic, state or activity that prohibits or detracts from safe, efficientnuclear plant operation or a condition that could credibly impact nuclearsafety, personnel safety, plant reliability or non-compliance with federal, state,
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1L-12-015Page 5 of 24or local regulations. Adverse conditions that are failures, malfunctions,deficiencies, deviations, defective hardware and non-conformances, orhuman performance, programmatic, organizational, or managementweaknesses that adversely affect Quality (Q), Augmented Quality (AQ), ornuclear safety-related equipment, programs, or processes, are consideredconditions adverse to quality. Adverse conditions include conditions adverseto quality, plant reliability issues, any concern that should be trended (e.g.,personnel contamination, personnel safety,,and unexpected plant equipmentfailures), and conditions that have significance within a regulatory context.FENOC does not differentiate whether an adverse condition is aging relatedwhen initiating a condition report; adverse conditions resulting from agingrelated issues are .documented in the Corrective Action Program in the samemanner as any other adverse condition. The Corrective Action Program,therefore, is considered the primary source of plant-specific operatingexperience since adverse conditions, including aging-related adverseconditions, are documented in the program database." Plant maintenance activities -maintenance activities -such as preventivemaintenance tasks, inspections, examinations, surveillances, tests oranalyses, are sources of plant-specific operating experience, includingpotential aging issues. Adverse conditions identified during the performanceof these types of activities are documented in the Corrective Action Program.* Plant Operator tours -Operator tours are performed routinely throughout thestation and provide the opportunity to identify adverse system, structure orcomponent conditions, including aging issues. Adverse conditions identifiedduring the performance of Plant Operator tours are documented in theCorrective Action Program." System Engineer walkdowns -Periodic System Engineer walkdownsperformed as part of the System Performance Monitoring Program providethe opportunity to identify adverse system, structure or componentconditions, including aging issues. Adverse conditions identified during thecourse of System Engineer walkdowns are documented in the CorrectiveAction Program.* Aging Management Program activities -preventive maintenance tasks,inspections or examinations performed at the direction of aging managementprograms are sources of plant-specific operating experience, including agingissues. Additionally, trending of program inspection or examination resultsand effectiveness reviews required by select aging management programsmay also provide a source of plant-specific operating experience, includingaging issues. Adverse conditions identified during the performance of thesetypes of activities are documented in the Corrective Action Program.
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1L-12-015Page 6 of 24-(b) Indicate whether plant-specific and industry operating experience is onlyconsidered from a prescribed list of sources. 'If only- prescribed sources areconsidered, provide a justification as to why it is unnecessary to considerother sources.FENOC does not consider plant-specific and industry operating experience onlyfrom a prescribed list of sources.For plant-specific operating experience, FENOC considers input from many sources,such as those described in the response to question (a), above. Adverse conditionsidentified from these sources are documented in the FENOC Corrective ActionProgram. The Corrective Action Program procedure does not include a prescribedlist of sources from which to identify adverse conditions; rather, the procedureprovides examples of potential adverse conditions so that plant personnel willidentify plant issues from a wide range of sources.Although the Operating Experience Program does not restrict the use of othersources of operating experience, it does require- screening of industry operatingexperience from a list of prescribed sources at a minimum. The list of prescribedoperating experience sources includes:* Institute of Nuclear Power Operations (INPO) Event Reports;* INPO Significant Operating Experience Reports (SOERs);* INPO Operating Experience Reports;* Nuclear Regulatory Commission (NRC) Information Notices (INs);* NRC Regulatory Issue Summaries (RISs); and,* NRC Regulatory changes.The Operating Experience Program procedure also allows for processing operatingexperience from additional sources not listed above on a case-by-case basis.Other programs and processes, such as those provided in the examples identifiedbelow, are also available to FENOC for obtaining industry operating experience andaging-related information:The Continuous Equipment Performance Improvement process (deriveddirectly from INPO's AP-913, "Equipment Reliability Process Description")requires periodic updates to the process based on evaluation of plant-specificand industry operating experience from sources such as new vendorrecommendations, Electric Power Research Institute (EPRI) preventivemaintenance templates, the INPO Equipment Performance and InformationExchange (EPIX) system, or aging studies.
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1L-12-015Page 7 of 24The Materials Degradation Management Program is an industry initiative toadopt Nuclear Energy Institute (NEI) guidance for managementof materialissues. The Program is applicable to materials of construction within theReactor Coolant System Pressure Boundary, although-it can be applied tomaterials of construction within other systems. The Materials DegradationManagement Program ensures, among other issues, that materialS operatingexperience is shared among utilities.Industry owners groups, vendors, and EPRI committees, industrymanaged groups and Users groups share information that can help identifyaging concerns.Industry benchmarking lessons learned.(c) Indicate whether guidance documents are considered as a source of operatingexperience information. If they are considered as a potential source, provide aplan for considering the content of guidance documents, such as the GALLReport, as operating experience applicable to aging management.FENOC considers NRC Information Notices, Regulatory Issue Summaries, andRegulatory changes as sources of operating experience. However, FENOC does notconsider NRC guidance documents as sources of operating experience information.Guidance documents, such as the NRC NUREG-Series publications, are reports orbrochures on regulatory decisions, results of research, results of incidentinvestigations, and other technical and administrative information. These documentsare not sources of operating experience; however, the NRC may use NUREGs tocompile and report the results of operating experience research from industrysources. For example, new License Renewal Interim Staff Guidance documents orrevisions to NUREG-1800, "Standard Review Plan for Review of License RenewalApplications for Nuclear Power Plants," and NUREG-1 801, "Generic Aging LessonsLearned (GALL) Report," provide, in addition to license renewal guidance, the resultsof research on previously-identified industry operating experience and the associatedlessons-learned, and are historical documents by the time they are issued.FENOC is actively involved with numerous industry users groups and committees,which regularly share information on new industry operating experience. Changesto guidance documents are often identified and discussed at these industrymeetings, allowing FENOC to be informed of the issuance of new or revisedguidance documents.While the Operating Experience Program procedure does not include NRC guidancedocuments on the list of operating experience sources to be reviewed, the proceduredoes include Regulatory Issue Summaries, which can be used by the NRC to inform
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1L-12-015Page 8 of 24licensees of changes to guidance documents. For example, FENOC was madeaware of Revision 2 to NUREG-1800 and 1801 through the Nuclear Energy Institute(NEI) License Renewal Working Groups and through Regulatory Issue Summary2011-05, "Information on Revision 2 to the Generic Aging Lessons Learned Reportfor License Renewal of Nuclear Power Plants." Regulatory Issue Summary 2011-05was screened by FENOC using the Operating Experience Program.(d) Describe how operating experience issues will be identified and categorizedas related to aging.Plant-specific aging-related operating experience issues documented in the FENOCCorrective Action Program are processed and investigated in the same manner asother adverse conditions. However, the Corrective Action Program NuclearOperating Procedure is planned to be revised to require that a condition reportinvestigation of an aging-related issue for structures and passive componentsinclude consideration of the affected structure or component, material, environment,aging effect, aging mechanism, and aging management program, with feedback tothe affected aging management program owner for consideration of the impact toaging management program effectiveness.Industry aging-related operating experience items are received and screened nodifferently than other industry operating experience items received through existingoperating experience pipelines. Aging-related operating experience items are nottypically received with an "aging" flag or designator; rather, the operatingexperience items identify a condition or event experienced at another station. Forexample, corrosion-related or cracking failures may not necessarily be designatedas "aging-related."The Fleet Operating Experience Program Manager and Station OperatingExperience Coordinators review INPO's Nuclear Network and incomingcorrespondence for new external operating experience items. A screening process isperformed for the external operating experience items to determine the susceptibilityof a similar and unacceptable event or condition occurring at FENOC. If theoperating experience item is screened as not applicable (NA) to FENOC, no furtheraction is taken. An operating experience item that is determined to be applicable toFENOC is further screened as follows:If the operating experience item represents a potential, operability orreportability concern, the operating crew Shift Manager or Shift Engineer isnotified and a condition report is'written to document the issue in theCorrective Action Program.* If the operating experience item represents an adverse condition, a conditionreport is written to document the issue in the Corrective Action Program.
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1L-12-015Page 9 of 24If the operating experience item is identified for an "Evaluation RequiredReview," a formal evaluation with supervisory or management oversightis initiated.If the operating experience item requires further screening, it is submitted toan Operating Experience Coordinator peer group, which may include subjectmatter experts as needed, for a more detailed review to determinesusceptibility to FENOC stations.If the operating experience item is screened as applicable, but may notrequire a document evaluation, the item is considered "Information Only," andis-listed in the Weekly Operating Experience Summary.o. The Section Operating Experience Coordinator is responsible forreviewing the Weekly Operating Experience Summary and determiningwhether any of the items require an evaluation, enlisting subject matterexperts as necessary to assist in the determination.** These Information Only operating experience reports are alsocommunicated and distributed to personnel via the Weekly OperatingExperience Summary. No specific evaluation is required for operatingexperience items identified as Information Only reviews. However, ifthe reviewer determines that further evaluation is required, thereviewer should initiate an Evaluation Required Review by contactingthe Fleet or Station Operating Experience Coordinator.The Operating Experience Program Nuclear Operating Business Practice is plannedto be revised to require that an Evaluation Required Review of aging-relatedoperating experience issues for structures and passive components includeconsideration of the affected structure or component, material, environment, agingeffect, aging mechanism, and aging management program, with feedback to theaffected aging management program owner for consideration of the impact to agingmanagement program effectiveness.Additionally, the Corrective Action Review Board Nuclear Operating BusinessPractice is planned to be revised to ensure that the Corrective Action Review Boardquestions whether aging was considered for condition report investigations that arereviewed by the Board.See the enhancements descriptions at the end of this response.
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1L-12-015Page 10 of 24(e) Describe the training requirements on aging issues for those plantpersonnel responsible for screening, evaluating, and submitting operatingexperience items.The training requirements for personnel responsible for screening, evaluating andsubmitting (to the industry) operating experience items consist of completion ofposition-specific document and procedure reviews, proficiency demonstrations andinterviews that provide personnel with the ability to independently perform theactivity. Mentoring is used during the training process to provide direction, coaching,and oversight by a recognized technical expert who is qualified in the specific taskrequirements to ensure job performance requirements are understood andcompetency is achieved.A training "needs analysis" is planned to determine and document recommendedenhancements to the training requirements for those personnel responsible forscreening, evaluating and submitting (to the industry) aging-related operatingexperience items.See the enhancements descriptions at the end of this response.(f) Describe how evaluations of operating experience issues related to aging willconsider the following:* systems, structures, or components* materials* environments* aging effect* aging mechanisms* AMPsFor plant-specific aging-related operating experience issues, the Corrective ActionProgram Nuclear Operating Procedure is planned to be revised to require that acondition report investigation of an aging-related issue for structures and passivecomponents include consideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging management program, withfeedback to the affected aging management program owner for consideration of theimpact to aging management program effectiveness.For industry aging-related operating experience items, the Operating ExperienceProgram Nuclear Operating Business Practice is planned to be revised to requirethat an Evaluation Required Review of aging-related operating experience issues for
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1L-12-015Page 11 of 24structures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and agingmanagement program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(g) Describe how the results of the AMP inspections, tests, analyses, etc., will beconsidered as operating experience.Degraded or non-conforming conditions or results that do not meet the definedacceptance criteria identified during aging management program inspections, tests,analyses or other program activities are considered adverse conditions and aredocumented in condition reports using the FENOC Corrective Action Program.These results and the associated condition report are considered operatingexperience for the affected aging management program.The applicable site or fleet Management Review Board is responsible to reviewcondition reports to identify internal operating experience that has the potential to beshared with the other FENOC sites and the industry. The following attributes areused to make the determination:* Important to nuclear, public, radiological, and personnel safety* Important to power generation capability* Events with important generic implications* Events for which an investigation was performed and the lessons learnedwould be beneficial to know about if the event had occurred at another station* Events required to be reported by INPO Event Report (IER) Level 1 orSOER recommendationsThe results of aging management program inspections, tests, analyses or otherprogram activities that meet the defined acceptance criteria are also consideredoperating experience for the affected aging management program. These results areused as feedback to the aging management program for trending purposes, and forevaluation to determine, based on the component, material, environment and agingeffect combinations managed, whether the frequency of future inspections needs tobe adjusted or new inspections established.The Corrective Action Program Nuclear Operating Procedure is planned to berevised to require that a condition report investigation of an aging-related issue for,structures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and aging
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1L-12-015Page 12 of 24management program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(h) Describe the operating experience evaluation records with respect to what isconsidered for aging. Indicate whether these records are maintained inauditable and retrievable form.Operating experience evaluation records are processed using existing plantprocedures. Plant-specific operating experience is documented in condition reportsand processed using the FENOC Corrective Action Program. Aging issues thatwould be captured in the Corrective Action Program include adverse conditionssuch as the following:0 aging-related degradation that adversely affects or threatens plant systems,structures or components;* adverse aging-related trends identified in system health reports or agingmanagement programs;0 aging management program weaknesses or programs identified asineffective; or,* results that do not meet the acceptance criteria from inspections,examinations, tests, analyses, or other activities performed under an agingmanagement program.Condition reports for adverse conditions and related documents captured inthe Corrective Action Program database are quality records and are auditableand retrievable.In a similar manner, industry operating experience sources are monitored andoperating experience, including aging-related operating experience, is entered intothe -Operating Experience Program database. External operating experience itemsare exposed to a screening process that provides for three possible outcomes:, Not applicable (NA) to FENOC, and no further action is taken;* Applicable-to FENOC, but may not require a documented evaluation, and aretherefore considered "Information Only;" or,* Applicable-to FENOC and require an evaluation ("Evaluation Required Review").Operating experience reports that are screened as "Information Only" are unlikely toaffect nuclear or personnel safety, plant reliability or availability, or are unlikely toneed follow-up actions. Information Only operating experience items are.
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1L-12-015Page 13 of 24communicated and distributed to personnel in a weekly operating experiencesummary. If a reviewer later determines that further evaluation is required, thereviewer should initiate an "Evaluation Required Review" by contacting the Fleet orStation Operating Experience Coordinator.An Evaluation Required Review is performed for operating experience reports thatare determined to require a FENOC or site evaluation, or when the initial screeningidentifies the need for further evaluation and/or actions to address the operatingexperience issue. The degree of rigor involved in evaluating operating experienceinformation should be based on the significance of the issue identified. Documentscaptured in the Operating Experience Program database are retrievable.(i) Provide details on the operating experience evaluation schedules and justifywhy they provide for timely evaluations. Also, describe how the relativesignificance of operating experience items is determined so that the reviewscan be prioritized appropriately.Plant-specific operating experience is documented in condition reports andprocessed using the FENOC Corrective Action Program. Supervisor reviews ofcondition reports are expected to be performed as soon as possible, normally withinone business day of initiation of the condition report. The Management ReviewBoard, which consists of a collegial review and concurrence by the Managers from arepresentative cross-section of plant disciplines, reviews new condition reports dailyand establishes the due dates for condition report evaluations. Due dates are setbased on the safety significance of the issue; the standard due date for conditionreport evaluations is 30 days. For plant-specific operating experience that should beshared with the industry, the goal is to issue the report to the industry within 50 daysof the origination date of the associated condition report.Industry operating experience is documented in the Operating Experience databaseand processed using the FENOC Operating Experience Program. If the operatingexperience issue is screened such that an Evaluation Required Review is assignedand no operability, reportability or adverse conditions are identified, then time isavailable to properly evaluate the issue. For operating experience Evaluation RequiredReviews for INPO Level 1 and 2 Event Reports, INPO is to be informed of the FENOCaction plan within 150 days. For Evaluation Required Reviews for other operatingexperience items, the evaluation due date is initially established as 150 days.The determinations of the relative significance and the resulting determination oftimeliness for the evaluation of operating experience items are subjective, but arebased on the plant knowledge and experience of the reviewers. Considerationsinclude items such as the real or potential impact to nuclear safety, generation risk,plant-operation, systems, structures or components, programs, potential operability
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1L-12-015Page 14 of 24or reportability concerns, or whether an adverse condition exists. The real orpotential impacts may not be obvious, and research may need to be performed todetermine whether the station has similar systems, structures or components,materials, or conditions such that the station would be affected.(j) Justify why the corrective action program has an appropriate threshold forcapturing issues concerning aging.The FENOC Corrective Action Program has an appropriate threshold for capturingissues concerning aging because the program requires that adverse conditions,regardless of their nature (including aging-related degradation), are documented in acondition report in the Corrective Action Program. The training provided to plantpersonnel coupled with procedural guidance and supervisory and managementoversight ensures that adverse conditions, including issues concerning aging, areappropriately captured and evaluated. Periodic regulatory and utility-basedassessments and effectiveness reviews of the Corrective Action Program are usedto confirm that adverse conditions are appropriately addressed.(k) Describe the criteria for considering when AMPs should be modified or newAMPs developed due to operating experience. Also, describe the process forimplementing changes to the AMPs or for implementing new AMPs; describehow these changes are implemented in a timely manner.The FENOC Corrective Action Program requires development of actions that areeffective at addressing conditions adverse to quality. Examples of effectivecorrection actions that address aging may include modification of existing ordevelopment of new aging management programs. Adverse aging conditions maybe identified via plant-specific or industry operating experience that indicate one ormore of the following:* a new aging effect is identified;" the applicable aging effect or aging mechanism for a given combination ofsystem, structure or component, material, environment, aging effect andaging mechanism is not being effectively managed;" adverse aging-related trends are identified in system health reports or agingmanagement programs;" aging management program weaknesses are identified or programs areidentified as ineffective; or,
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1L-12-015Page 15 of 24, results are obtained that do not meet the acceptance criteria from inspections,examinations, tests, analyses, or other activities performed under an agingmanagement program.Revision of existing or development of new aging management programs based onoperating experience evaluations would be performed through corrective actionsusing the Corrective Action Program, or by action items identified in the OperatingExperience Program database. Assigned program owners would develop revisionsto their assigned aging management program implementing procedures based onthe results of an evaluation of the operating experience issues. A regulatoryapplicability review and, as applicable, a 10 CFR 50.!59 review are performed toconfirm whether the change can be made without prior NRC approval, and whetherthe change impacts any information in the Davis-Besse Updated Safety AnalysisReport. New aging management program implementing procedures would bedeveloped based on the activities involved and the systems, structures orcomponents affected. The revised or new aging management program implementingprocedure would undergo an internal FENOC review and approval process. The duedates for corrective actions and Operating Experience Program action, items areclosely tracked and managed.For plant-specific aging-related operating experience issues, the Corrective ActionProgram Nuclear Operating Procedure is planned to be revised to require that acondition report investigation of an aging-related issue for structures and passivecomponents include consideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging management prograrfi, withfeedback to the affected aging management program owner for consideration of theimpact to aging management program effectiveness.For industry aging-related operating experience items, the Operating ExperienceProgram Nuclear Operating Business Practice is planned to be revised to requirethat an Evaluation Required Review of aging-related operating experience issues forstructures and passive components include consideration of the affected structure orcomponent, material, environment, aging effect, aging mechanism, and, agingmanagement program, with feedback to the affected aging management programowner for consideration of the impact to aging management program effectiveness.See the enhancements descriptions at the end of this response.(I) Provide criteria for reporting plant-specific operating experience onage-related degradation to the industry.Noteworthy plant-specific operating experience is shared with the other FENOCsites and the industry. The applicable site or fleet Management Review Board is
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1L-12-015Page 16 of 24responsible to identify internal operating experience that has the potential to beshared with the other FENOC sites and the industry. The following attributes areused to make the determination:0 Important to nuclear, public, radiological, and personnel safety* Important to power generation capability* Events with important generic implications* Events for which an investigation was performed and the lessons learnedwould be beneficial to know about if the event had occurred at another station* Events required to be reported by INPO Event Report (IER) Level 1 orSOER recommendationsThe Operating Experience Program procedure provides the guidance on sharinginternal operating experience.If enhancements are necessary, provide an implementation schedule forincorporating them into the existing programmatic operating experience reviewactivities.The following training activities and procedure changes will be completed on or beforeDecember 31, 2012:1. Perform a training needs'analysis to determine and-document recommendedenhancements to the training requirements for those plant personnel responsiblefor screening, evaluating and submitting (to the industry) aging-related operatingexperience items. Based on the results of the training needs analysis, identify theappropriate training materials.2. Revise Nuclear Operating Business Practice NOBP-LP-2100, "FENOCOperating Experience Process," to require that an Evaluation Required Reviewof aging-related operating experience issues for structures and passivecomponents includes:a. consideration of the material, environment, aging effect, agingmechanism, and aging management program for the affected structure orcomponent; and,b. a provision for feedback to the affected aging management programowner for consideration of the impact to aging management programeffectiveness.
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1L-12-015Page 17 of 243. Revise Nuclear Operating Procedure NOP-LP-2001, "Corrective ActionProgram," to require that a condition report investigation of an aging-related issuefor structures and passive components includes:a. consideration of the material, environment, aging effect, agingmechanism, and aging management program for the affected structure orcomponent; and,b. a provision for feedback to the affected aging management programowner for consideration of the impact to aging management programeffectiveness.4. Revise Nuclear Operating Business Practice NOBP-LP-2008, "FENOCCorrective Action Review Board," to ensure that the Corrective Action ReviewBoard questions whether aging was considered for condition report investigationsthat are reviewed by the Board.See Attachment 2 to this letter for the regulatory commitments.Question RAI B.1.4-3Background:In RAI B.1.4-1, the staff asked the applicant to provide, in accordance with10 CFR 54.21(d), a USAR supplement a summary description of theprogrammatic activities for the ongoing review of operating experience, asrequired by 10 CFR 54.21(d). By letter dated August 17, 2011, the applicantprovided this description:Existing FENOC processes require reviews of relevant site andindustry operating experience and periodic benchmarking toensure program enhancements are identified and implemented.Such ongoing reviews identify potential needs for agingmanagement program revisions to ensure their effectivenessthroughout the period of extended operation.Issue:As described above in RAI B.1.4-2, the applicant described generally how itintends to consider operating experience on an ongoing basis; however, it didnot provide specific information on how its operating experience review activitiesaddress issues related to aging. Similarly, the above entry for USAR supplement
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1.L-12-015Page 18 of 24also lacks details on how aging is considered in the ongoing operatingexperience reviews.Request:,Consistent with the response to RAI B.1.4-2, provide additional details in theUSAR supplement on how the ongoing operating experience review activitiesaddress issues specific to aging.RESPONSE RAI B.1.4-3FENOC replaces the operating experience summary description in License RenewalApplication (LRA) Section A. 1, "Summary Descriptions of Aging Management Programsand Activities," provided by letter dated August 17, 2011 (ML1 1231A966), with a newdescription that is consistent with the FENOC response to RAI B.1.4-2, above.See Enclosure A to this letter for the revision to the DBNPS LRA.Section 2.1Supplemental Question RAI 2.1-3 Abandoned EquipmentIn a supplemental response to RAI 2.1-3 titled, "SUPPLEMENTAL RESPONSE -Abandoned Equipment," submitted by letter dated September 16, 2011(ML11264A059), FENOC provided a plan to ensure abandoned equipment isidentified, isolated and drained, as follows:1. Determine the scope of abandoned equipment -includes review of Piping& Instrumentation Diagrams (P&IDs), plant walkdowns, and review of theShift Operations Management System (eSOMS) clearance database.2. Determine the status of abandoned equipment -includes review of systemstatus files and the eSOMS database for as-left valve positions, walkdownsto validate valve position status, and ultrasonic testing to confirm thatabandoned piping is drained.3. Place abandoned equipment in a configuration that will not impactsafety-related equipment -create and implement Operations EvolutionOrders to isolate and drain abandoned systems with fluids, and create and
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1L-12-015Page 19 of 24implement Document*Change Requests as necessary to correct theconfiguration of the plant as shown on plant drawings.SUPPLEMENTAL RESPONSE RAI 2.1-3 Abandoned EquipmentFENOC completed the actions to ensure abandoned equipment is identified, isolatedand drained as provided in the plan submitted by FENOC letter datedSeptember 16, 2011 (ML11264A059). Identification of the scope of abandonedequipment that could impact safety-related equipment was determined through a reviewof Piping & Instrumentation Diagrams (P&IDs), plant walkdowns, and review of the ShiftOperations Management System (eSOMS) clearance database.The status of the abandoned equipment that could impact safety-related equipment wasdetermined through review of system status files and the eSOMS database for as-leftvalve positions, walkdowns to validate valve position status, and ultrasonic testing toconfirm that abandoned piping is drained.Abandoned equipment that could impact safety-related equipment was verified to beisolated and drained with the exception of components associated with the ServiceWater System intake crib air bubbler compressors, and the Miscellaneous LiquidRadwaste System degasifier skid, miscellaneous waste evaporator skid, evaporatorstorage tank pumps, and primary water transfer pumps. The subject components areadded tothe scope of license renewal in accordance with 10 CFR 54.4(a)(2), as follows:* The Lake Erie elevation is such that it provides a nonisolable source of water tothe discharge piping associated with the abandoned intake crib air bubblercompressors. The components associated with these discharge lines containfluid, and are, therefore, added to the scope of license renewal. License renewalboundary drawing LR-M012E is revised to highlight these additional components.LRA Section 2.3.3.26, "Service Water System," is revised to add license renewalboundary drawing LR-M012E to the list of license renewal boundary drawings thatdepict the evaluation boundaries of the Service Water System. LRA Table 2.3.3-26, "Service Water System Components Subject to Aging Management Review,"is revised to identify a structural integrity function for the orifice component type.LRA Table 3.3.2-26, "Aging Management Review Results -Service WaterSystem," is revised to provide the updated aging management review results forthese additional components.The components associated with the abandoned degasifier skid, miscellaneouswaste evaporator skid, evaporator storage tank pumps, and primary watertransfer pumps contain fluid or are not sufficiently isolated to remain drained, andare, therefore, added to the scope of license renewal. New License renewalboundary drawings LR-M01OD Sheet 2, LR-M036C Sheet 2 and LR-M039BSheet 2 were created, and license renewal boundary drawings LR-M01OD
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1L-12-015Page 20 of 24Sheet 1, LR-M020B, LR-M036C Sheet 1, LR-M037C, LR-M037D, LR-M037E,LR-M037F, LR-M039A, LR-M039B Sheet 1 and LR-M042B are revised tohighlight the additional components. LRA Section 2.3.3.21, "Miscellaneous LiquidRadwaste System," is revised to add additional license renewal boundarydrawings to the list of boundary drawings that depict the evaluation boundaries ofthe system. LRA Table 2.3.3-21, "Miscellaneous Liquid Radwaste SystemComponents Subject to Aging Management Review," is revised to include thenew component types. LRA Table 3.3.2-21, "Aging Management Review Results-Miscellaneous Liquid Radwaste System," is revised to provide the updatedaging management review results for the additional components.See Enclosure A to this letter for the revision to the DBNPS LRA.See Enclosure B to this letter for the revision to the LRA Boundary Drawings.Section 3.1.2Supplemental Question RAI 3.1.2.2-2 Reactor Vessel Internals Aging ManagementThe NRC initiated a telephone conference call with FENOC on January 24, 2012,to discuss the pressurized water reactor (PWR) Reactor Vessel Internals Programand how FENOC plans to implement MRP-227-A, "Materials Reliability Program:Pressurized Water Reactor Internals Inspection and Evaluation Guidelines(MRP-227-A)." The staff wanted to know FENOC's plans to address anydifferences between MRP-227 and MRP-227-A.FENOC responded that a review of MRP-227-A dated December 2011 against theprevious version of MRP-227 dated December 2008 was in progress to identifyany changes needed to the Davis-Besse PWR Reactor Vessel Internals (RVI)Program or to the aging management review (AMR) results provided in LicenseRenewal Application (LRA) Table 3.1.2-2, "Aging Management Review Results -Reactor Vessel Internals." FENOC noted that, at a minimum, several changes arerequired to LRA Table 3.1.2-2.FENOC agreed to provide a response to address the differences betweenMRP-227-A and the previous version of MRP-227, and provide the necessaryupdates for the reactor vessel internals AMR results and the PWR Reactor VesselInternals (RVI) Program.
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1L-12-015Page 21 of 24The staff requested that FENOC be very clear as to what'has and what has notchanged, and to identify what was reviewed.SUPPLEMENTAL RESPONSE RAI 3.1.2.2-2 Reactor Vessel Internals Aging ManagementAging Management Review Results for the Reactor Vessel InternalsIn response to RAI 3.1.2.2-3 submitted by FENOC letter dated September 16,-2011(ML1 1264A059), LRA Table 3.1.2-2, "Aging Management Review Results -ReactorVessel Internals," was replaced in its entirety. These aging management review (AMR)results were based on Electric Power Research Institute (EPRI) Report 1016596,"Materials Reliability Program: Pressurized Water Reactor Internals Inspection andEvaluation Guidelines (MRP-227-Rev. 0)," dated December 2008. This report has sincebeen updated with the changes proposed by the NRC in the safety evaluation forMRP-227 as well as changes proposed by the EPRI Materials Reliability Program inresponse to NRC requests for additional information (RAIs) and issued as EPRI Report1022863, "Materials Reliability Program: Pressurized Water Reactor InternalsInspection and Evaluation Guidelines (MRP-227-A)," dated December 2011.FENOC completed a review of MRP-227-A dated December 2011 and NRC safetyevaluation, "Revision 1 to the Final Safety Evaluation of Electric Power Research Institute(EPRI Report), Materials Reliability Program (MRP) Report 1016596 (MRP-227),Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and EvaluationGuidelines" (TAC No. ME0680)," against the previous version of MRP-227 datedDecember 2008, and determined that changes are required to the AMR results providedin LRA Table 3.1.2-2. A summary of the required table changes are listed as follows:Row 4 -control rod guide tube (CRGT) spacer casting changed from an expansioncomponent with primary component link of core support shield (CSS) cast outletnozzles, CSS vent valve discs or in-core monitoring instrumentation (IMI) guidetube spiders to a primary component with no expansion components. This revisedclassification is consistent with Section 3.3.7 of the NRC safety evaluation.Row 10 -CSS cast outlet nozzles changed from a primary component withexpansion component of CRGT spacer casting to a 'no additional measures'component. This revised classification is consistent with Table 3-1 ofMRP-227-A. As provided in note 2 of the table, "Thermal Embrittlement (TE)revised from 'P' to 'A' based on review of material data (ferrite content) fromONS-3 and DB; final grouping accordingly changed from 'Primary' to 'NoAdditional Measures."'"Row 12 -As discussed in Section' 3.7 of the-safety evaluation, the CSS ventvalve disc was determined to be an active component and not subject to agingmanagement. Therefore, row 12 is changed to "Not used."
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1L-12-015Page 22 of 24" Row 13 -As discussed in Section 3.7 of the safety evaluation, the CSS ventvalve disc shaft was determined to be an active component and not subject toaging management. Therefore, row 13 is changed to "Not used."* Row 20 -In addition to the aging effects of cracking due to irradiation-assistedstress corrosion cracking (IASCC) and reduction in fracture toughness, the agingeffects of cracking due to fatigue, loss of material and loss of preload were addedfor the core barrel-to-former (CBF) bolts. These revised aging effects ateconsistent with Table 3-1 of MRP-227-A.* Row 21 -In addition to the aging effects of cracking due to IASCC and reductionin fracture toughness, the aging effects of cracking due to fatigue, loss of materialand loss of preload were added for the baffle-to-former (FB) bolts. These revisedaging effects are consistent with Table 3-1 of MRP-227-A.* Row 22 -In addition to the aging effects of cracking due to fatigue and reductionin fracture toughness, the aging effects of loss of material and loss of preloadwere added for the baffle-to-baffle (BB) bolts -internal. These revised agingeffects are consistent with Table 3-1 of MRP-227-A." Row 23 -In addition to the aging effects of cracking due to fatigue, cracking dueto IASCC and reduction in fracture toughness, the aging effects of loss of materialand loss of preload were added for the baffle-to-baffle (BB) bolts -external.These revised aging effects are consistent with Table 3-1 of MRP-227-A.* Row 42 -Since the CRGT spacer casting was changed to a primary component(see Row 4 above), it is deleted as an expansion component for the IMI guide tubespiders. This revised classification is consistent with Table 4-1 of MRP-227-A." Row 43 -Since the CRGT spacer casting was changed to a primary component(see Row 4 above), it is deleted as an expansion component for the IMI guidetube spider-to-lower grid rib section welds. This revised classification isconsistent with Table 4-1 of MRP-227-A." Plant-specific note 0114 -This note addressed the classification of the flowdistributor (FD) bolts and their locking devices but is no longer needed for thatpurpose since Table 4-1 of MRP-227-A now shows the component as a primarycomponent with expansion components of upper thermal shield (UTS) bolts andtheir locking devices, lower thermal shield (LTS) bolts and their locking devices,and surveillance specimen holder tube (SSHT) bolts and their locking devices.This revised classification is consistent with Section 4.1.3 of the NRC safetyevaluation. Plant-specific note 0114 is now used to provide clarification thatcomponents assigned to the no additional measures group were determined-tobe below the screening criteria for the applicable degradation mechanisms, orwere classified under this category due to Failure Modes, Effects, and Criticality
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1L-12-015Page 23 of 24Analyses (FMECA) and functionality analysis findings, and therefore, no furtheraction is required by MRP-227-A for managing the aging of these components.PWR Reactor Vessel Internals ProgramIn response to RAI B.2.32-1 submitted by FENOC letter dated September 16, 2011(ML1 1264A059), the Davis-Besse PWR Reactor Vessel Internals Program wasreplaced in its entirety. This revised program, provided in LRA Section B.2.32 and thecorresponding Updated Safety Analysis Report (USAR) Supplement (LRA SectionA.1.32), addressed each of the five plant-specific aging management programinformation requirements identified in Section 3.5.1 of the NRC safety evaluation, "FinalSafety Evaluation of EPRI Report, Materials Reliability Program Report 1016596(MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection andEvaluation Guidelines" (TAC No. ME0680)," dated June 22, 2011.FENOC has completed a review of the five plant-specific aging management programinformation requirements identified in Section 3.5.1 of the NRC safety evaluation,"Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRIReport), Materials Reliability Program (MRP)_Report 1016596 (MRP-227), Revision 0,"Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines"(TAC No. ME0680)," dated December 16, 2011, and determined that no changes arerequired to the Davis-Besse PWR Reactor Vessel Internals Program relative to addressingeach of the five plant-specific aging management program information requirements.As noted in the above discussion (see "Row 13" bulleted item) that addressed therequired changes to LRA Table 3.1.2-2, the CSS vent valve disc shaft was determinedto be an active component and not subject to aging management. Therefore, LRASections A.1.32 and B.2.32, both titled "PWR Reactor Vessel Internals Program," arerevised to delete any discussion of the vent valve disc shaft.In addition, the 5th paragraph of the "Detection of Aging Effects" program element forLRA Section B.2.32 was provided to address the flow distributor (FD) bolts classificationand examination method/coverage/frequency. Since MRP-227-A has incorporated thechanges outlined in Section 4.1.3 of the safety evaluation relative to the flow distributor(FD) bolts, the 5th paragraph is no longer needed and is therefore deleted.Also, LRA Sections A.1.32 and B.2.32 are also revised to incorporate MRP-227-Aversus MRP-227, Revision 0, as the program implementation document.See Enclosure A to this letter for the revision to the DBNPS LRA.
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1L-12-015Page 24 of 24Section 4.2Supplemental Question Section 4.2 Reactor Vessel Neutron EmbrittlementThe NRC initiated a telephone conference call with FENOC on February 9, 2012, todiscuss time-limited aging analyses (TLAAs) associated with Section 4.2 of theLicense Renewal Application (LRA). Two of the topics discussed during thetelephone conference are as follows.In LRA Section 4.2.1, the neutron fluence analysis for 52 EFPY (60-years ofoperation) was dispositioned as not a time-limited aging analysis (TLAA). Thestaff's position is that neutron fluence is a TLAA. FENOC agreed to change theLRA to disposition the neutron fluence analysis as a TLAA using10 CFR 54.21(c)(1)(ii).In LRA Section 4.2.6 including the associated USAR Supplement Section A.2.2.6,discussion was provided to address the future replacement of the reactor vesselclosure head. FENOC agreed to change the LRA to acknowledge that the reactorvessel closure head was replaced in the fall of 2011.SUPPLEMENTAL RESPONSE Section 4.2 Reactor Vessel Neutron EmbrittlementAs provided in LRA Section 4.2.1, "Neutron Fluence," a neutron fluence analysis validfor 52 effective full power years (EFPY) has been prepared for the reactor vesselbeltline materials to bound the projected value of 50.3 EFPY for 60-years of operation.LRA Section 4.2.1, including the associated USAR Supplement Section A.2.2.1,"Neutron Fluence," and LRA Table 4.1-1, "Time-Limited Aging Analyses," are revised todisposition the neutron fluence analysis as a time-limited aging analysis (TLAA) inaccordance with 10 CFR 54.21 (c)(1)(ii).LRA Section 4.2.6, "intergranular Separation (Underclad Cracking)," including theassociated USAR Supplement Section A.2.2.6, "lntergranular Separation -UndercladCracking," addressed the future replacement of the reactor vessel closure head/headflange schedule for the fall of 2011. Also, the operating experience program element forLRA Section B.2.29, "Nickel-Alloy Reactor Vessel Closure Head Nozzles Program,"addressed the future replacement of the head. The reactor vessel closure head/headflange was replaced in the fall of year 2011. Therefore, LRA Sections 4.2.6, A.2.2.6 andB.2.29 are revised to provide this updated status for the reactor vessel closure head.See Enclosure A to this letter for the revision to the DBNPS LRA.
Attachment
2L-12-015Regulatory Commitment ListPage 1 of 2The following list identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)in this document. Any other actions discussed in the submittal represent intended orplanned actions by FENOC; they are described only as information and are notRegulatory Commitments. Please notify Mr. Clifford I. Custer, Project Manager -FleetLicense Renewal, at (724) 682-7139 of any questions regarding this document orassociated Regulatory Commitments.Regulatory CommitmentDue Date1. Perform a training needs analysis to determine December 31, 2012and document recommended enhancements tothe training requirements for those plantpersonnel responsible for screening, evaluatingand submitting (to the industry) aging-relatedoperating experience items. Based on the resultsof the training needs analysis, identify theappropriate training materials.2. Revise Nuclear Operating Business Practice December 31, 2012NOBP-LP-2100, "FENOC Operating ExperienceProcess," to require that an Evaluation RequiredReview of aging-related operating experienceissues for structures and passive componentsincludes:a. consideration of the material, environment,aging effect, aging mechanism, and agingmanagement program for the affectedstructure or component; and,b. a provision for feedback to the affectedaging management program owner forconsideration of the impact to agingmanagement program effectiveness.
Attachment
2L-12-015Page 2 of 2Regulatory CommitmentDue Date-I-3. Revise Nuclear Operating ProcedureNOP-LP-2001, "Corrective Action Program," torequire that a condition report investigation of anaging-related issue for structures and passivecomponents includes:a. consideration of the material, environment,aging effect, aging mechanism, and agingmanagement program for the affectedstructure or component; and,b. a provision for feedback to the affectedaging management program owner forconsideration of the impact to agingmanagement program effectiveness.December 31, 2012 "4. Revise Nuclear Operating Business Practice December 31, 2012NOBP-LP-2008, "FENOC Corrective ActionReview Board," to ensure that the CorrectiveAction Review Board-questions whether agingwas considered for condition report investigationsthat are reviewed by the Board.
Enclosure
ADavis-Besse Nuclear Power Station, Unit No. I (DBNPS)Letter L-12-015Amendment No. 24 to theDBNPS License Renewal ArplicationPage 1 of 51License Renewal ApplicationSections AffectedSection 2.3.3.21Table 2.3.3-21Section 2.3.3.26Table 2.3.3-26Table 3.1.2-2Table 3.1 Plant-Specific NotesTable 3.3.2-21Table .3.3.2-26Section 4.2.1.3Section 4.2.6Section A. 1Section A.1.32Section A.2.2.1Section A.2.2.6Section B.2.29Section B.2.32Table 4.1-1The 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 fined-ou and added text underlined.
Enclosure
AL-12-015Page 2 of 51Affected LRA Section LRA Page No. Affected Paragraph and Sentence2.3.3.21 Page 2.3-113 License Renewal Drawings,9 additional drawingsrIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Section2.3.3.21, "Miscellaneous Liquid Radwaste System," is revised to add nineadditional license renewal boundary drawings to the list of "License RenewalDrawings," to read as follows:License Renewal DrawingsThe following license renewal drawings depict the evaluation boundaries for thesystem components within the scope of license renewal:LR-MOIOD Sheet 1, LR-MOIOD Sheet 2, LR-MO2OB, LR-M031A, LR-M033A,LR-M036A, LR-M036C Sheet 1, LR-M036C Sheet 2, LR-M037C, LR-M037D,LR-M037E, LR-M037F, LR-M037G, LR-M039A, LR-M039B Sheet 1,LR-M039B Sheet 2, LR-M042B, LR-M045, LR-M046, LR-M281N13
Enclosure
AL-12-015Page 3 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceTable 2.3.3-21 Page 2.3-114 12 new rowsIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Table2.3.3-21, "Miscellaneous Liquid Radwaste System Components Subject to AgingManagement Review," is revised to include 12 new rows for new componenttypes, and reads as follows:Component Type Intended Function(as defined in Table 2.0-1)Heat exchanger (shell) -Degasifier Heat Exchangers Structural integrityHeat exchanger (shell) -Waste Evaporator Heat, Structural integritExchangersPump casing -Anti Foam Tank Pump (DB-P55 WM) Structural integrityPump casing -Degasifier Discharge Pump Structural integrityPump casing -Evaporator Storage Tank Pumps Structural integrity(DB-P53-1 & 2)Pump casing -Primary Water Transfer Pumps Structural integrity(DB-P41-1, 2)Pump casing -Waste Evaporator Bottoms Pump Structural-integrityPump casing -Waste. Evaporator Distillate Pump Structural integrityPump casing -Waste Evaporator Vacuum Pump Structural integrityTank -Anti Foam Tank (DB-T30) Structural integrityTank -Degasifier (DB-S3) Structural integrityTank -Waste Evaporator Concentrator (DB-S2) Structural integrity
Enclosure
AL-12-015Page 4 of 51Affected LRA SectionLRA Page No.Page 2.3-128Affected Paragraph and SentenceLicense Renewal Drawings,1 additional drawing2.3.3.26In response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Section2.3.3.26, "Service Water System," is revised to add one additional license renewalboundary drawing to the list of "License Renewal Drawings," to read as follows:License Renewal DrawingsThe following license renewal drawings depict the evaluation boundaries for thesystem components within the scope of license renewal:LR-MOO6D, LR-MO12E, LR-M036A, LR-M036B, LR-M041A, LR-M041B, LR-M041CAffected LRA SectionLRA Page No.Page 2.3-130Affected Paragraph and SentenceTable 2.3.3-26Intended FunctionIn response to Supplemental RAI 2.1-3 Abandoned Equipment, LRA Table2.3.3-26, "Service Water System Components Subject to Aging ManagementReview," is revised to include the "structural integrity" intended function for theorifice component type, and reads as follows:Component Type Intended Function(as defined in Table 2.0-1)Pressure boundaryOrifice Structural integrityThrottling.
Enclosure
AL-12-015Page 5 of 51Affected LRA SectionTable 3.1.2-2Table 3.1.2 Plant-Specific NotesLRA Page No.Pages 3.1-60thru 3.1-121Page 3.1-187Affected Paragraph and Sentence27 Rows, various columnsNote 0114In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals Aging Management, 27 rows of LRATable 3.1.2-2, "Aging Management Review Results -Reactor Vessel Internals," and Plant-Specific Note 0114,previously replaced in their entirety by FENOC letter dated September 16, 2011 (ML1 1264A059), are revisedto read as follows:MIo0y A- IOU uuweIs-tL- PWR ReactorPlenum Cover Borated Reactor Vessel Internals None ABottomNickel Coolant with Cracking -SCC (IV.B4.RP- None 0Flange Welds Support Alloy 0114(no additional Neutron Fluence PWR Water 236)measures component) Chemistry
Enclosure
AL-12-015Page 6 of 514(expansion componentwith p~mriry Gomponentlink of CSS Cast OutletNozz~es, CSS VentValve~ DiBscs orIM!(primary componentwith no expansioncomponents)SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracturetoughnessPWR ReactorVessel InternalsIV. B4-4(IV.B4.RP-242)3.1.1-80ACRGT Rod GuideTubes Stainless Borated Reactor Loss of PWR Reactor None A5 Support Coolant with Loss NoneSupot olat it material -wear Vessel Internals 23V) B41P14on(no additional Steel Neutron Fluence 237) 0114measures component)CRGT Rod GuideSectors Stainless Borated Reactor Loss of PWR Reactor None A(n diinlSupport Sanes Coolant with(I.4R- Nn6 Suppona Steel Neutron Fluence material -wear Vessel Internals (IV.B4.RP- None 0114meaddition ent)237)___measures component) _________________________________________
Enclosure
AL-12-015Page 7 of 51Upper Core Barrel(UCB) Bolts (originalbolts) and their lockingdevicesBolt: Cracking -SCCPWVK ReactorVessel InternalsIV.1B4-20(IV.B4.RP-248)3.1.1-37A0444PWR WaterChemistry8(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,.and SSHT Bolts andtheir locking devices)SupportStainlessSteelBorated ReactorCoolant withNeutron FluenceLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV. B4-01(IV.B4.RP-243)C3.1.1-22____________________ _________________ ___________________________ __________________________ +/- ___________________________ I _____________________ L _______________ I __________
Enclosure
AL-12-015Page 8 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRo~wNo.9Component TypeIntendedFunction(s)MaterialEnvironmentAging EffectRequiringManagementAging 'ManagementProgramNUREG-1801Rev. I Item(Rev. 2 Item)Table IItemNotesF t I F I t I IUpper' Core Barrel(UCB) Bolts(replacement bolts) andtheir locking devices(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)CSS Cast OutletNozzles(primar-y cOMPonentcomponents 414 fWlOW,:CRGT Spacer Casting)(no additionalmeasures component)CSS Vent Valve Discs(primary ,OMPnentwith e ,PWR ReactorVessel InternalsBolt: Cracking -SCCIV.1B4-20(IV.B4.RP-248)3.1.1-37A0444SupportBolt:NickelAlloyLockingDevices:StainlessSteelBorated ReactorCoolant withNeutron FluenceBolt:Cumulative IV. B4-37 3 1.1-05 Afatigue damage (IV. B4. R-53)-fatiguePWR WaterChemistryLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV. B4-01(IV.B4.RP-243)3.1.1-22C10SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracture.toughnessPWR ReactorVessel InternalsIV. B4-21(IV.B4.RP-253)3.1.1-80A011412Naomt Us ed Not Used&49petF46W~GR*nGO;astAustenoStainlessStee4Berated Reacte,GeelaR*-tNeutron FluenceReduetien infraetU~etoughnessPKVFR ReactorVessRel InternalI-V -84-24(~L84Rpz344. 0AI I_____ I ____________________ I __________ 4 ___________________________ 1 __________________________ 4 ... ______________
Enclosure
AL-12-015Page 9 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type ntn Material Environment Requiring Management Rev. I Item NotesManagement Program (Rev. 2 Item)CSS Vent Valv-e DiscShaf tB4 1...ompoen Berated Reacto Re& in PW .I-3 4.RPQ13 ,M no e n suppe--o4 Vessel , ,,; 344 2 AGOwit noexpnsio L)p# Neutron Fluenc toughness OV-.B4.RP-compenents) Not Used \Core Barre[AssemblPWR ReactorAlloy X-750 Vessel Internals None. ACore Barrel-to-Former Borated Reactor Cracking -SCC (IV.B4.RP- None 011415 Plate Dowel Support Nickel PWR Water 236)15 PaeDwlSpot Alloy Coolant with Chemistry(n o additional Neutron Fluence Reduction in Nonemeasures component) fracture PWR Reactor (IV.B4.RP- None AVessel Internals. " 0114toughness 237)Alloy X-750 PWR ReactorDowel-to-Core Barrel Borated Reactor Vessel Internals None A16 Cylinder Fillet Welds Support Nickel Coolant with Cracking -SCC (IV.B4.RP- None(no additional Alloy Neutron Fluence PWR Water 236) 0114measures component) ChemistryThermal Shield Upper Cracking -Restraint Cap Borated ReactoIr ueNone.17 Screws (Not Exposed) Support Stess Coolant with Loss of (IV.B4.RP- None(no additional Neutron Fluence material -wear Vessel Internals 237) 0114measures component) Loss of preload
Enclosure
AL-12-015Page 10 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1o. Component Material Environment Requiring Management Rev. 1 Item NotesManagement Program (Rev. 2 Item) ItemIV.B4-07(IV.B4.RP-PWR Reactor 244);Cracking -Vessel Internals (IV.B4.RP-Core Barrel-to-Former IASCC, Fatigue 238) 3.1.1-30 A(CBF) Bolts Borated Reactor PWR Water None20 (expansion component Support Stainless Coolant with Chemistry (IV.B4.RP-with primary component Steel Neutron Fluencelink of FB Bolts) Reduction infracture IV.1B4-01toughness PWR Reactor (IV.B4.RP-Loss of Vessel Internals 243); 3.1.1-22 Amaterial -wear (IV. B4. R P-239)Loss of preload
Enclosure
AL-12-015Page 11 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging_ NUREG-1801 Table INo. Component Type Intended Material Environment Requiring Management Rev. I Item NotesFunction(s) Management Program (Rev. 2 Item) ItemIV.B4-07PWR Reactor (IV.B4.RP-Cracking -Vessel Internals 241) 3.1.1-30 ABaffle-to-Former (FB) IASCC, FatiQue PWR Water None 3Bolts___Borated Reactor Chemistry (IV.B4.RP-21 (primary component Support Stainless Colntwt 37L5121 with expansion Steel Coolant withcomponents of BB Neutron Fluence fractureBolts and CBF Bolts) toughness PWR Reactor IV.B4-01Loss of Vessel Internals (IV.B4.RP- 3.1.1-22 Amaterial -wear 240)Loss of preloadPWR ReactorVessel Internals NoneCracking -(IV.B4.RP- None ABaffle-to-Baffle (BB) fatigue PWR Water 375)Bolts -internal Borated Reactor Chemistry22 (expansion component Support Stainless Coolant with Reduction inwith primary component Steel Neutron Fluence fracturelink of FB Bolts) toughness PWR Reactor IV. B4-01Loss of Vessel Internals (IV.B4.RP- 3.1.1-22 Amaterial -wear 243)Loss of preload
Enclosure
AL-12-015Page' 12 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow .Intended Aging Effect Aging NUREG-1801 Table 1Row Component Type Intended Material Environment Requiring Management Rev. I Item NotesNo. Function(s) Management Program (Rev. 2 Item) ItemPWR ReactorCracking -Vessel Internals Noneratinge (IV.B4.RP- None Afatigue PWR Water 375)ChemistryBaffle-to-Baffle (RB) PWR Reactor IV.1B4-07Bolts -external Boae Vessel Internals (IV.B4.RP-Stainless Borated Reactor Cracking- 244); 3.1.1-30 A23 (expansion component Support Steel Coolant with IASCC PWR Water (IV.B4.RP-with primary component Neutron Fluence Chemistry 238)link of FB Bolts) Reduction infracture IV. B4-01toughness PWR Reactor (IV.B4.RP-Loss of Vessel Internals 243);material -wear (IV.B4.RP-Loss of preload 239)Lower Core Barrel PWR Reactor(LCB) Bolts (original) Bolt: Cracking -Vessel Internals IV.R4-13 Aand their locking SCC (IV.B4.RP-. 3.1.1-37 0-44devices PWR Water 247)(primary component Borated Reactor Chemistry26 with expansion Support Ste Coolant withcomponents of UTS Steel Neutron Fluence LockingBolts and their locking Device: PWR Reactor IV. B4-01devices, LTS Bolts and Loss of Vessel Internals (IV.B4.RPJ 3.1.1-22 Ctheir locking devices, material -wear 243)and SSHT Bolts andtheir locking devices)
Enclosure
AL-12-015Page 13 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type Function(s) Material Environment Requiring Management Rev. I Item NotesNo. Fnis _______ Management Program (Rev. 2 Item) Item27Lower Core Barrel(LCB) Bolts(replacement) and theirlocking devices(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)Bolt: Cracking -SCCPWR ReactorVessel InternalsIV.B4-13(IV.B4.RP-247)3.1.1-37SupportBolt:NickelAlloyLockingDevices:StainlessSteelBorated ReactorCoolant withNeutron FluenceBolt:Cumulative .IV.B4-37 3.1.1-05 Afatigue damage (IV.B4.R-53)-fatiguePWR WaterChemistryA0444LockingDevice:.Loss ofmaterial -wearPWR ReactorVessel InternalsIV.1B4-01(IV.B4.RP-243)3.1.1-22C30Alloy X-750 Dowel-to-Upper Grid Rib SectionBottom Flange Welds(noadditionalmeasures component)SupportNickelAlloyBorated ReactorCoolant withNeutron FluenceCracking -SCCPWR ReactorVessel InternalsPWR WaterChemistryNone(IV.B4.RP-236)NoneA011435Alloy X-750 Dowel-to-Lower GridShell Forging Welds(no additionalmeasures component)SupportNickelAlloyBorated ReactorCoolant withNeutron FluenceCracking -SCCPWR ReactorVessel InternalsNone(IV.B4.RP-236)NoneA0114PWR WaterChemistry
Enclosure
AL-12-015Page 14 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table 1No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item NotesManagement Program (Rev. 2 Item)PWR ReactorVessel Internals NoneAAlloy X-750 Dowel-to- Cracking -ALower Grid Rib Borated Reactor SCC, IASCC (IV.B4.RP- NoneNickel PWR Water 236) 011436 Section Welds Support Alloy Coolant withChemistry(no additional Neutron Fluence Chemistrymeasures component) Reduction in PWR Reactor None Afracture Vessel Internals (IV.B4.RP- None 0114toughness 237)Lower Grid Rib-to-Shell Cracking -Forging Cap Borated Reactor fatigue NoneScrews Stainless PWR Reactor f37 Support Steel Coolant with Loss of Vessel Internals o (IV.B4.RP- None 0114(no additional Neutron Fluence material -wear 237) Noe-11measures component) Loss of preloadLower Grid Support Cracking -Post Pipe Cap Borated Reactor fatigue None38 Screws Support Stainless Coolant with Loss of V esl ter (IV.B4.RP- None A(no additional Steel Neutron Fluence material -wear 237)measures component) Loss of preload
Enclosure
AL-12-015Page 15 of 51ti-ow uisirnutor (I-u)Bolts and their lockingdevicesPWi< ieactorVessel InternalsBolt: CrackingSCCIV.B4-25(IV.B4.RP-256)A ,3.1.1-37 A044-440(primary componentwith expansioncomponents of UTSBolts and their lockingdevices, LTS Bolts andtheir locking devices,and SSHT Bolts andtheir locking devices)SupportStainlessSteelPWR WaterChemistryBorated ReactorCoolant withNeutron FluenceLockingDevice:Loss ofmaterial -wearPWR ReactorVessel InternalsIV.B4-01(IV.B4.RP-243)3.1.1-22 1 CAlloy X-750,Dowel-to- PWR ReactorFlow Distributor Borated Reactor Vessel Internals None41 Flange Welds Support Alloy Coolant with Cracking -SCC (IV.B4.RP- None 0(no additional .Neutron Fluence PWR Water 236) 0114measures component) Chemistry
Enclosure
AL-12-015Page 16 of 5142IMI (5uiae I uve spicters(primary componentwith expansioncomponents of-GRGTSpacor- Casting an~Lower Fuel AssemblySupport Pads: Pad,Pad-to-Rib SectionWeld, Cap Screw andassociated LockingWeld, Alloy X-750Dowel and Alloy X-750Dowel Locking Weld)SupportCastAusteniticStainlessSteelBorated ReactorCoolant withNeutron FluenceReduction infracturetoughnessPWR ReactorVessel InternalsIV. B4-28(IV.B4.RP-258)3.1.1-80A_____________ ____________ J. __________________ -I _________________ .1 __________________ .1. ______________ 1 _________ ~I. _______
Enclosure
AL-12-015Page 17 of 51Table 3.1.2-2 Aging Management Review Results -Reactor Vessel InternalsRow Intended Aging Effect Aging NUREG-1801 Table INo. Component Type Fntinds) Material Environment Requiring Management Rev. I Item NotesNo. Function(s) Management Program (Rev. 2 Item) "ItemIMI Guide Tube Spider-to-Lower Grid RibSection Welds(primary componentwith expansioncomponents of-CRGTSpa.er. Casting an Stainless Borated Reactor Reduction in PWR Reactor IV.1B4-3143 Lower Fuel Assembly Support Steel Coolant with fracture Vessel Internals (IV.B4.RP- 3.1-.1-22 ASupport Pads: Pad, Neutron Fluence toughness 259)Pad-to-Rib SectionWeld, Cap Screw andassociated LockingWeld, Alloy X- 750Dowel and Alloy X-750Dowel Locking Weld)Plant-Specific Notes:'0114 Flow distribto ~F~blswr esined as a primary component per- MRP--227-, Rev. 0 as amended by the safet evaluation.Due to thiS haG , t--FDbolts are not anepnso omponent for- the UCB and LOB bolts.Components assigned to the "No Additional Measures" group were determined to be below the screening criteria for theaiplicable degradation mechanisms, or were classified under this category due to Failure Modes, Effects, and Criticality Analyses(FMECA) and functionality analysis findings, and therefore, no further action is required by MRP-227-A for managing the aging ofthese components.
Enclosure
AL-12-015Page 18 of 51Affected LRA SectionLRA Page No..Page 3.3-411thru 3.3-417Affected Paragraph and SentenceTable 3.3.2-2136 New Rows; and,Rows 55, 60, 65, 70, 75, 79, & 84, "Notes" columnIn response to Supplemental RAI 2.1-3 Abandoned Equipment, 36 new rows are added to LRATable 3.3.2-21, "Aging Management Review Results -Miscellaneous Liquid Radwaste System." Additionally,during the review, it was identified that generic Note "A" was assigned in the "Notes" column for rows 55, 60,65, 70, 75, 79, and 84, but should have been Note "C". LRA Table 3.3.2-21 is revised to include new rows andcorrect assigned notes, as follows:Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemTAging Effect NUREG-Row Component Intended Material Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItemHeatExchanger Collection, Drainage,(shell) -Structural Stainless Raw water Loss of and Treatment VII.C15 3.3.1-79 E-Degasifier integrity Steel (Internal) material Components VHeat InspectionExchangersHeatExchanger Air with(shell) -Structural Stainless borated water None None VIIJ-16 3.3.1-99 C-Degasifier inte~grity Steel leakageHeat (External)Exchangers
Enclosure
AL-12-015Page 19 of 51Table 3.3.2-21 Aging Management Review Results -.Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table NotesNo. Type Function(s) Program Volume 2 itemManagement Item .IHeatExchanger Air-indoor(shell) -] Structural Stainless Aridoesher itrity steel uncontrolled None None VII.J-15 3.3.1-94 C-- Detasfir Steel (External)HeatExchangersHeatExchanger(shaell) -Collection, Drainage,(shell)_- Structural Stainless Raw water Loss of and Treatment-- Waste ..Sel (nenl aeilCmoet VII.CI-15 3.3.1-79 EEvaporator integrit Steel (Internal) material ComponenHeat InspectionExchangersHeatExchanger. Air with(shell) Structural Stainless borated waterWaste None None VII. J- 16 3.3.1-99 CEvaporator inert Steel leakage .Nn oe___ ___Heat o (External)Exchangers
Enclosure
AL-12-015Page 20 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 NotesNo. Type Function(s) Program Volume 2 ItemManagement I ItemHeatExchanger(shell) Structural Stainless Air-indoorWaste uncontrolled None None VII.J-15 3.3.1-94 C.Evaporator i Steel (External)HeatExchangersPump casing Collection, Drainage.-Anti Foam Structural Stainless Raw water Loss of and Treatment-- Tank Pump VII._ C1-5 _.31-9_TnB- integrity Steel (Internal) material Components VICI-15 3.3.1-79 EP55 WM) InspectionPump casing Air with-Anti Foam ArwtTank Fuoa Structural Stainless borated water-- Tank Pump itgiy Sel eageNone None VII.J-16 3.3.1-99 A(DB- Steel leakaeP55 WM) (External)Pump casing-Anti Foam Air-indooStructural Stainless Air-indoorTank Pump Integrity Steel uncontrolled None None VII.J-15 3.3.1-94 A(DB- (External)P55 WM)
Enclosure
AL-12-0.15Page 21 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemdAging Effect NUREG-Row Component Intended Material Environment Requiring Agn aaeet 10, Tbe1 NotesNo. Type Function(s) M Program Volume 2 ItemMaeral Enirnmn Rqurig gigManagement18, Tal NoeItemPump casing Collection, Drainage,-Degasifier Structural Stainless Raw water Loss of and Treatment VIICI-15 3.3.1-79 E-Dischare integrity Steel (Internal) material ComponentsPUMP InspectionPump casing Air with-Degasifier Structural Stainless borated water None None VIIJ-16 3.3.1-99 ADischarge integrit Steel leakagePump (External)Pump casing Air-indoor-Degasifier Structural Stainless Aridouncontrolled None None VII.J-15 3.3.1-94 ADischarge integrity Steel (External)PumpPump Casing Collection, Drainage,-Evaporator-- Storage Tank Structural Stainless Raw water Loss of and Treatment VII.C1-15 3.3.1-79 EPumPs (DB- integrit Steel (Internal) material Components,P53-1,. 2) InspectionPump Casing Air with-Evaporato Structural Stainless borated water-- Storage Tank Strut Stail boate None None VII. J- 16 3.3.1-99 APumps (DB- i Steel leakageP53-1, 2) (External)
Enclosure
AL-12-015Page 22 of 51Table 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 ItemItemPump Casing-EvaporaTa Structural Stainless Air-indoor-Storage Tank Strur Stanle uncontrolled None None VII.J-15 3.3.1-94 APumps (DB- n Steel (ExternalP53-1, 2)Pump casing Collection, Drainage,-Primary Structural Stainless Raw water Loss of and Treatment-- Water VI1_C1-15 3.3.1-79Transfer inte-grit Steel (Internal) material Componentsu.Trane Inspection* PumpPump casing Air with-Primary Structural Stainless borated water-- Water SNone None VII.J-16 3.3.1-99 ATransfer ,negt Steel leakagePum.n (External)Pump casing-Primary Structural Stainless Air-indoorWater uncontrolled None None VII.J-15 3.3.1-94 ATransfer integrity Steel (External)PumPPump casing Collection, Drainage,-Waste Structural Stainless Raw water Loss of and Treatment VEvaporator integrit Steel (Internal) material Components VII-IS 3.3.1-79 EBottoms InspectionPump I I____
Enclosure
AL-12-015Page 23 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAgin EfectNUREG-Row Component Aging Management 1801, Table tggfcNMaterial Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItemPump casing Air with-Waste Structural Stainless borated water-- Evaporator integrit Steel leakage None None VII.J-16 3.3.1-99 ABottoms (External)'PumPPump casing-Waste. Air-indoo-- EWasorator Structural Stainless Air-indoor-- EvaporatorS uncontrolled None None VII.J-15 3.3.1-94 ABottoms n Steel (External)PumpPump casin Collection, Drainage,-Waste Structural Stainless Raw water " Loss of and Treatment-E rato integrit Steel (Internal) material ComponentsDistillate InspetoPump IpectionPump casing Air with-Waste Structural Stainless borated water None None VII.J-16 3.3.1-99 ADEvsporator integrity Steel leakageNDistillate (External)Pump casing-Waste Structural Stainless Air-indoor-- Evaporator i t Steel uncontrolled None None VII.J-15 3.3.1-94 ADistillate (External)Pump_
Enclosure
AL-12-015Page 24 of 51Table 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 Management Program Volume 2 ItemItemPump casingPm Wasig Collection, Drainage,Structural Stainless Raw water Loss of and TreatmentEvaporator integrity Steel (Internal) material Components VII.C1-15 3.3.1-79 EVacuum IsetoPumpPump casing Air with-Waste Structural Stainless borated waterEvaporator inte.grit Steel leakage None None VII.J-16 3.3.1-99 AVacuum inertSte leagPumum (External)Pump casing-Waste Structural Stainless Air-indoorEvaporator integrit Steel uncontrolled None None VII.J-15 3.3.1-94 A.Vacuum (External)PumpTank -Anti Collection, Drainage,TankoamAntik Structural Stainless Raw water Loss of and TreatmentFoam Tank integrity Steel (Internal) material Components VII.C-15 3.3.1-79 E(DB-T30) InspectionSTank -Anti Air withFoam Tank Structural Stainless borated water None None VII.J-16 3.3.1-99 C(DBT30) integrity Steel leaka-ge(External)
Enclosure
AL-12-015Page 25 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemeAging Effect TNUREG-No TnMing Effect Aging Management 1801, Table 1No. Type Function(s) aing Program Volume 2 Item NotesM ItemTank -Anti structural Stainless Air-indoor--_Foam Tank Strutur Steele uncontrolled None None VII.J-15 3.3.1-94 C(DB-T30) i Steel (External)Tank- Collection, Drainagqe,Degasifier Structural Stainless Raw water Loss of and Treatment V11.C1-5 3.3.1-79 E-(DB-S3) integrit Steel (Internal) material Components VInspectionTank- Air withDegasifier Structural Stainless borated water None None VIIJ-16 3.3.1-99 C-(DB-S3) integrity Steel leakage(External)Tank -Degasifier Structural Stainless Air-indoor(DB-S3) ntegrity Steel uncontrolled None None VII.J-15 3.3.1-94 C(External)Tank- Collection, Drainage,Eartr Structural Stainless Raw water Loss of and TreatmentConcentrator i Steel (Internal) material Components(DB-S2) Inspection
Enclosure
AL-12-015Page 26 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemAging Effect NUREG-Row Component Intended MaterialAging Management 1801, Table I NotesNo. Type Function(s) Mong Program Volume 2 Item1Management ItemTank -Air withWaste Structural Stainless borated waterCEvaporator intevrit Steel leakage None None VII.J-16 3.3.1-99 CConcentrator 6 Stel erakae,(DB-(External)Tank -Waste Air-indoorWaste Structural Stainless AridoEvaporator Strut Stail uncontrolled None None VII.J-15 3.3.1-94 CConcentrator i Steel (External)(DB-S2Tank -Structural Stainless Air-indoor A55 DWDT 1-1 integrity Steel uncontrolled None None VII.J-15 3.3.1-94 C(DB-T27) (External)Tank -Air-indoorDWDT 1-1 Structural Stainless Ai-ndo60 uncontrolled None None VII.J-15 3.3.1-94 CHold-up Tank integrity Steel (External) C(DB-T161)Tank-Miscellaneous Air-indoor65 Liquid Waste itrity Steel uncontrolled None None VII.J-15 3.3.1-94 CMonitor Tank integrity Steel (External) C(DB-T29)
Enclosure
AL-12-015Page 27 of 51Table 3.3.2-21 Aging Management Review Results -Miscellaneous Liquid Radwaste SystemdAging Effect NUREG-Row Component Intended Material Environment Requiring NotesNo. Type Function(s) Management Program Volume 2 ItemItem _Tank -Miscellaneo structural Stainless Air-indoor70 Waste Drain Stegructu Stale uncontrolled None None VII.J-15 3.3.1-94 CTank i(External)(DB-T26)Tank -Miscellaneous Ari d oWaste Structural Stainless Air-indoorA75 Waste Strity Stess uncontrolled None None VII.J-15 3.3.1-94 AEvaporator integrity Steel (External)Storage Tank(DB-T28)Tank -Radwaste structural Stainless Air-indoor A79 Vessel integrity Steel uncontrolled None None VII.J-15 3.3.1-94Skid Vessel i(External)(1 through 5)Tank -Waste84 Polishing Structural Stainless uncontrolled None None VII.J-15 3.3.1-94 ADemineralizer integrity Steel unotrolle C(DB-T125) (External)
Enclosure
AL-12-015Page 28 of 51Affected LRA SectionLRA Page No.Page 3.3-482Affected Paragraph and SentenceTable 3.3.2-269 New RowsIn response to Supplemental RAI 2.1-3 Abandoned Equipment, nine new rows are added to LRATable 3.3.2-26, "Aging Management Review Results -Service Water System," to read as follows:Table 3.3.2-26 Aging Management Review Results -Service Water SystemAging Effect NUREG-Row Component Intended Material Environment Requiring -Aging Management 1801, Table I NotesNo. Type Function(s) Management Program Volume 2 Item NItemStructural Stainless Raw water Loss of Open-Cycle CoolingOrifice ntegrity Steel (Internal) material Water VII.Cl-15 3.3.1-79 BStructural Stainless Air-indoorSuuOrifice Steel uncontrolled None None VII.J-15 3.3.1-94 A-- reintegrit Steel (External)Structural Stainless Condensation Loss of External SurfacesOrifice integrity Steel (External) material Monitoring VI-I 3.3.1-27 EStructural Stainless Raw water Loss of Open-Cycle Cooling VII. C1-15 3.3.1-79 B-- _ inte-grity Steel (Internal) material Water
Enclosure
AL-12-015Page 29 of 51Table 3.3.2-26 Aging Management Review Results -Service Water SystemAging Effect NUREG-Row Type Fntinds) Material Environment Requiring Aging Management 1801, Table INo. Type Function(s) Management Program Volume 2 Item NotesManagmentItemStructural Stainless Air-indoorStegructu Stanl uncontrolled None None VII.J-15 3.3.1-94 A-ntegrt Steel (Extemal)Structural Stainless Condensation Loss of External Surfacesinte-grity Steel (External) material Monitoring VII.Fl1-1 3.3.1-27 EStructural Stainless Raw water Loss of Open-Cycle Coolinq VII.C1-15 3.3.1-79 BValve Body integrity Steel (Internal) material WaterStructural Stainless Air-indoorSuuValve Body Steel uncontrolled None None VII.J-15 3.3.1-94 A(External)-- Valve Body Structural Stainless Condensation Loss of External Surfaces VII.FI-1 33.1-27 EBody integrity Steel (External) material Monitoring
Enclosure
AL-12-015Page 30 of 51Affected LRA SectionLRA Page No.Affected Paragraph and SentenceTable 4.1-1Page 4.1-3"Neutron Fluence" rowIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the "Neutron Fluence" row of LRA Table 4.1-1, "Time-LimitedAging Analyses," is revised as follows:Table 4.1-1 Time-Limited Aging Analyses54.21 (c)(1) LRAResults of TLAA Evaluation by Category Paragraph LRAParagraph SectionReactor Vessel Neutron Fluence 4.2Neutron Fluence N1 t .-1.T,, 4.2.120Affected LRA SectionLRA Page No.Affected Paragraph and Sentence4.2.1.3Page 4.2-3New 4th paragraph; and,"Disposition"In response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, LRA Section 4.2.1.3, "Beltline Evaluation," is revised to include anew fourth paragraph and a revised "Disposition" statement, to read as follows:Neutron fluence analysis valid for 52 EFPY have been prepared for the reactorvessel beltline materials to bound the proiected value of 50.3 EFPY for 60-yearsof operation. Therefore, the neutron fluence analysis has been proiected to theend of the period of extended operation.Disposition;. Not a TLAA Neutron fluence is an assumption uei .n various neutron embrittlement TLAA&eva'uated be/ow.Disposition: 10 CFR 54.21(c)(1)(ii) The neutron fluence analysis has beenproiected to the end of the period ofextended operation.
Enclosure
AL-12-015Page 31 of 51Affected LRA Section LRA Page No. Affected Paragraph and Sentence4.2.6 Page 4.2-14 5th paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the fifth paragraph on page 4.2-14 of LRA Section 4.2.6,"lntergranular Separation (Underclad Cracking)," is revised to read as follows:As proevidend in Confiratory' Acton Letter-, Number 3 10 001, FENOC havoluntarily comm~ittedf to shutdowný the Davis Besse plant no later- than October-1L,2011, and replace the RV closure head. Therefor~e, the cUrrent head (purchasedfrom the Midland Plant andf installed during the cycle 13 refuelhing outage) is nofconsidered in the -,nderclad cracking evaluation. The replacement RV closureheadlhead fl#ange, to be installed during the October- 2011 outage, was fabricated,usi .ng SA 508 Class 3 m~aterial, whic~h is not susceptible to interg:anUlaseparatien&- The reactor vessel closure head/head flan-ge was replaced in theFall of year 2011. This replacement head was fabricated using SA-508 Class 3material, which is not susceptible to the subiect intergranular separations.Therefore, this replacement closure head/head flange is not considered in theunderclad cracking evaluation.
Enclosure
AL-12-015Page 32 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1 Page A-9 3rd paragraphIn response to RAI B.1.4-3, the third paragraph of LRA Section A.1, "SummaryDescriptions of Aging Management Programs and Activities," previously addedby FENOC letter dated August 17, 2011 (ML11231A966), is replaced in itsentirety to read as follows:Existing FE-=NQG processes reqire rv of relevant site and industoperating exeinc n proi benchmarking to ensure proqgramea.ncrme. nt -,rz .. .ident.if..d and implemented. Such nGoi.ng rviews idenpotential need fo r- agn aagement program rew .sions to ensure thei;offoctivenes thrtoughout the period of extendfed opeFratonOperating ExperienceThe intent of evaluatinq and incorporating operating experience lessons-learned,including aging-related lessons-learned, is prevention. The lessons-learned areused to improve plant operation, equipment material condition, and agingmanagement to minimize equipment degradation and prevent loss of equipmentintended functions. The FENOC Operating Experience Program processes andprocedures for the ongoing review of operating experience include the followinqattributes:* Personnel responsible for screening, evaluating and submitting (to theindustry) aging-related operating experience items are qualified for thetask.* While the programs and procedures may specify reviews of certainsources of information, such as NRC generic communications andInstitute of Nuclear Power Operations reports, they allow for any potentialsource of relevant plant-specific or industry operating experienceinformation.The processes are adequate so as to not preclude the consideration ofoperating experience related to.-aging management. The processes allowfor appropriately gathering information on structures and passivecomponents within the scope of license renewal, their materials,environments, aging effects, and aging mechanisms, and the aging
Enclosure
AL-12-015Page 33 of 51management programs credited for managing the effects of aging,including the activities under these programs (e.g., inspection methods,preventive actions or evaluation techniques).Plant-specific operating experience, including aging-related operatingexperience, is documented in condition reports and processed using theFENOC Corrective Action Program. Condition -reports for adverseconditions and related documents captured in the Corrective ActionProgram database are quality records and are auditable and retrievable.Industry operating experience, including aging-related operatingexperience, is entered into the Operating Experience Program databaseand screened for applicability to FENOC. Documents captured in theOperating Experience Program database are retrievable.Evaluations of internal and external aging-related operating experienceissues associated with structures and passive components includeconsideration of the affected structure or component, material,environment, aging effect, aging mechanism, and aging managementprogram, with feedback to the affected aging management program ownerfor consideration of the impact to aging management programeffectiveness.Aging management program owners review data, collected by programactivities, use the Corrective Action Program to document adverseconditions to ensure they will be addressed and corrected, maintainrequired records for the program, maintain the program current, andimplement revisions as needed based on program results and internal orexternal operating experience evaluations. Revision of existing ordevelopment of new aging management programs based on operatingexperience evaluations'is performed through corrective actions using theCorrective Action Program, or by action items identified in the OperatingExperience Program database.Noteworthy plant-specific aging-related operating experience is sharedwith the other FENOC sites and the industry. The Operating ExperienceProgram procedure provides guidance on sharing internal operatingexperience.
Enclosure
AL-12-015Page 34 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.1.32 Page A-21 Paragraphs 1, 2, 3, 7 and 12 (last)In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals AgingManagement, the 1st, 2nd, 3rd, 7th and 12th paragraphs of LRA Section A.1.32,"PWR Reactor Vessel Internals Program," previously replaced in its entirety byFENOC letter dated September 16, 2011 (ML11264A059), are revised to readas follows (affected sentences highlighted for clarity):A.1.32 PWR REACTOR VESSEL INTERNALS PROGRAMThe PWR Reactor Vessel Internals Program relies on implementation of theElectric Power Research Institute (EPRI) Report No.10169 1022863,"Materials Reliability Program: Pressurized Water Reactor Internals Inspectionand Evaluation Guidelines (MRP-227-)," and EPRI Report No. 1016609,"Materials Reliability Program: Inspection Standard for PWR Internals(MRP-228)," to manage the aging effects on the reactor vessel internal (RVI)components.This program is used to manage the effects of age-related degradationmechanisms that are applicable in general to the PWR RVI components atDavis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effectsinclude (a) various forms of cracking, including stress corrosion cracking (SCC),which also encompasses primary water stress corrosion cracking (PWSCC),irradiation-assisted stress corrosion cracking (IASCC), or cracking due tofatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracturetoughness due to either thermal aging or neutron irradiation embrittlement; and(d) loss of preload due to thermal and irradiation-enhanced stress relaxation orcreep. In addition, the program includes management of the time-limited aginganalysis (TLAA) identified in License Renewal Application (LRA) Section A.2.2.7for reduction in fracture toughness of the reactor vessel internals. This TLAA willbe managed in accordance with the implementation of the MRP-227 guidelines,ass almendod by the A MR 227 _sfoty .va.at"on,- including all activities associatedwith Davis-Besse's responses to plant-specific action items identified inSection 4.2 of the safety evaluation.The program applies the guidance in MRP-227, 0, as am.ended by thefor inspecting, evaluating, and, if applicable, dispositioningnon-conforming RVI components at Davis-Besse. The program conforms to thedefinition of a sampling-based condition monitoring program, as defined by theBranch Technical Position RSLB-1, with periodic examinations and otherinspections of highly-affected internals locations. These examinations providereasonable assurance that the effects of age-related degradation mechanisms
Enclosure
AL-12-015Page 35 of 51will be managed during the period of extended operation. The program includesexpanding periodic examinations and other inspections if the extent of thedegradation effects exceeds the expected levels.The MRP-227 guidance for selecting RVI components for inclusion in theinspection sample is based on a four-step ranking process. Through this process,the reactor internals were assigned to one of the following four groups: Primary,Expansion, Existing Programs, and No Additional Measures components.Definitions of each group are provided in GALL Chapter IX.B.The result of this four-step sample selection process is a set of Primary InternalsComponent locations for each of the three plant designs (Westinghouse,Combustion Engineering and Babcock & Wilcox) that are expected to show theleading indications of the degradation effects, with another set of ExpansionInternals Component locations that are specified to expand the sample shouldthe indications be more severe than anticipated. The degradation effects in athird set of internals locations are deemed to be adequately managed by ExistingPrograms. A fourth set of internals locations are deemed to require no additionalmeasures. As a result, the program typically identifies 5 to 15 percent of the RVIlocations as Primary Component locations for inspections, with another 7 to10 percent of the RVI locations to be inspected as Expansion Components, aswarranted by the evaluation of the inspection results. Another 5 to 15 percent ofthe internals locations are covered by Existing Programs, with the remainderrequiring no additional measures. This process thus uses appropriate componentfunctionality criteria, age-related degradation susceptibility criteria, and failureconsequence criteria to identify the components that will be inspected under theprogram in a manner that conforms to the sampling criteria for sampling-basedcondition monitoring programs in Section A.1.2.3.4 of NRC Branch PositionRLSB-1. Consequently, the sample selection process is adequate to assure thatthe intended function(s) of the PWR reactor internal components are maintainedduring the period of extended operation.No existing generic industry programs contain the specificity considered sufficientfor monitoring the aging effects addressed by the MRP-227 guidelines for B&Wplants. Therefore, no components for B&W plants were placed into the ExistingPrograms group.MRP-227 I&E guidelines require a visual (VT-3) examination of the core supportshield (CSS) vent valve retaining for every 10 year InserviceInspection Interval. In addition, Davis-Besse Technical Specification 5.5.4requires testing of the CSS vent valves every 24 months to verify by visualinspection that the valve body and valve disc exhibit no abnormal degradation,verify the valve is not stuck in an open position, and verify by manual actuationthat the valve is fully open when a force of 5400 lbs is applied vertically upward.
Enclosure
AL-12-015Page 36 of 51The technical specification inspection will continue to be performed at theprescribed frequency of 24 months. The MRP-227 required visual (VT-3)examination will also be performed at the prescribed frequency of every 10 yearInservice Inspection Interval.The program's use of visual examination methods in MRP-227 for detection ofrelevant conditions (and the absence of relevant conditions as a visualexamination acceptance criterion) is consistent with the ASME Code, Section XIrules for visual examination. However, the program's adoption of the MRP-227guidance for visual examinations goes beyond the ASME Code,Section XI visualexamination criteria because additional guidance is incorporated into MRP-227 toclarify how the particular visual examination methods will be used to detectrelevant conditions and describes in more detail how the visual techniques relateto the specific RVI components and how to detect their applicable age-relateddegradation effects.The technical basis for detecting relevant conditions using volumetric ultrasonictesting (UT) inspection techniques can be found in MRP-228, where the review ofexisting bolting UT examination technical justifications has demonstrated theindication detection capability of at least two vendors, and where vendortechnical justification is a requirement prior to any additional boltingexaminations. Specifically, the capability of program's UT volumetric methods todetect loss of integrity of PWR internals bolts, pins, and fasteners, such asbaffle-former bolting in B&W and Westinghouse units, has been welldemonstrated by operating experience. In addition, the program's adoption of theMRP-227 guidance and process incorporates the UT criteria in MRP-228, whichcalls for the technical justifications that are needed for volumetric examinationmethod demonstrations, required by the ASME Code,Section V.The program also includes future industry operating experience as incorporatedin periodic revisions to MRP-227. The. program thus provides reasonableassurance for the long-term integrity and safe operation of reactor internals in allcommercial operating U.S. PWR nuclear power plants.Age-related degradation in the reactor internals is managed through anintegrated program. Specific features of the integrated program are listed in thefollowing ten program elements. Degradation due to changes in materialproperties (e.g., loss of fracture toughness) was considered in the determinationof inspection recommendations and is managed by the requirement to useappropriately degraded properties in the evaluation of identified defects. Theintegrated program is implemented by the applicant through an inspection plan.The Davis-Besse PWR Reactor Vessel Internals Program will address allplant-specific action items applicable to Davis-Besse that are established in
Enclosure
AL-12-015Page 37 of 51Section 4.2 of the safety evaluation for MRP-227. In addition, a plant-specificinspection plan for ensuring the implementation of MRP-227 program guidelines,a. amended by the safoty 9a-uatiop for. and Davis-Besse's responsesto the plant-specific action items, as identified in Section 4.2 of the safetyevaluation for MR 22.", will be submitted for NRC review and approval.Affected LRA Section LRA Page No.Affected Para-graph and SentenceA.2.2.1Pages A-30and A-31Ist paragraph; andNew last paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 1st paragraph of LRA Section A.2.2.1, "Neutron Fluence," isdeleted, and a new last paragraph is added to the Section on LRA page A-31,as follows:1st Para-graphNeutroen flence is not a T-14, it is a timo limited assumptin used in theevaluation of neutron embrittlement TLAAs.New Last ParagraphA neutron fluence analysis valid for 52 EFPY has been prepared for the reactorvessel beltline materials to bound the proiected value of 50.3 EFPY for60 years of operation. Therefore, the neutron fluence analysis has beenproiected to the end of the period of extended operation in accordance with10 CFR 54.21(c)(1)(ii).
Enclosure
AL-12-015Page 38 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceA.2.2.6 Page A-34 3rd paragraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 3rd paragraph of LRA Section A.2.2.6, "Neutron Fluence," isrevised to read as follows:[Pro posed for this sc.tibn, pendnIg of Action Ltte:C41Ain 2o 10 001 conmmitments related to theg ,olcroto h ai ashead in 2011.] The reactor vessel closure head/head flange wasreplaced in the fall of year 2011. This replacement head was fabricated usingSA-508 Class 3 material, which is not susceptible to the subject intergranularseparations. Therefore, this replacement closure head/head flange is notconsidered in the underclad cracking evaluation.
Enclosure
AL-12-015Page 39 of 51Affected LRA Section LRA Page No. Affected Paragraph and SentenceB.2.29 Page B-119 2nd ParagraphIn response to Supplemental Question Section 4.2 Reactor Vessel NeutronEmbrittlement, the 2nd paragraph of the "Operating Experience" subsection ofLRA Section B.2.29, "Nickel-Alloy Reactor Vessel Closure Head NozzlesProgram," on LRA page B-119, is revised as follows:In March 2010, ultrasonic examinations of the control rod drive mechanismnozzles constructed of Alloy 600 material identified flaws on multiple nozzles.Active leakage was identified on one nozzle. The direct cause was PrimaryWater Stress Corrosion Cracking. The reactor vessel closure head had been inoperation approximately six years. An inside diameter temper bead half-nozzleweld repair was utilized. Post-repair inspections were completed with acceptableresults. As pro vided in Action Numbe ^h r: -3 10- 060 4, MaFk A.Satorius (NR1) to a-ry S. Allen dated 66 23 22010, hhavoluntaril committed to shutdown the Davis B esse plant no later- than October- 1,20114, and replace the reactor pressuq~re_ 14esse90 head- With one man1ufacturFed usingmaterials rosistant to PWSCC. The reactor vessel closure head was replacedwith a new head in the fall of year 2011. The CRDM nozzles for the new headwere fabricated using Alloy 690 material that is less susceptible to PWSCC.
Enclosure
AL-12-015Page 40 of 51Affected LRA Section LRA Page No. Affected Para-graph and SentenceB.2.32 Pages B-129 Various paragraphsthru B-133In response to Supplemental RAI 3.1.2.2-2 Reactor Vessel Internals AgingManagement, various paragraphs of LRA Section B.2.32, "PWR Reactor VesselInternals Program," previously replaced in its entirety by FENOC letter datedSeptember 16, 2011 (ML1 1264A059), are revised to read as follows (affectedsentences highlighted for clarity):B.2.32 PWR REACTOR VESSEL INTERNALS PROGRAMProgram DescriptionThe PWR Reactor Vessel Internals Program relies on implementation of theElectric Power Research Institute (EPRI) Report No.10659 1022863,"Materials Reliability Program: Pressurized Water Reactor Internals Inspectionand Evaluation Guidelines (MRP-227-.A)," and EPRI Report No. 1016609,"Materials Reliability Program: Inspection Standard for PWR Internals(MRP-228)," to manage the aging effects on the reactor vessel internal (RVI)components.This program is used to manage the effects of age-related degradationmechanisms that are applicable in general to the PWR RVI components atDavis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effectsinclude (a) various forms of cracking, including stress corrosion cracking (SCC),which also encompasses primary water stress corrosion cracking (PWSCC),irradiation-assisted stress corrosion cracking (IASCC), or cracking due tofatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracturetoughness due to either thermal aging or neutron irradiation embrittlement; and(d) loss of preload due to thermal and irradiation-enhanced stress relaxation orcreep. In addition, the program includes management of the time-limited aginganalysis (TLAA) identified in License Renewal Application (LRA) Section 4.2.7 forreduction in fracture toughness of the reactor vessel internals. This TLAA will bemanaged in accordance with the implementation of the MRP-227 guidelinesr-asa.meonedad by tho MRP 227- safty including all activities associatedwith Davis-Besse's responses to plant-specific action items identified inSection 4.2 of the safety evaluation.The program applies the guidance in MRP-227, Rev. 0, as am.en ed by th'for inspecting, evaluating, and, if applicable, dispositioningnon-conforming RVI components at Davis-Besse. The program conforms to thedefinition of a sampling-based condition monitoring program, as defined by the
Enclosure
AL-12-015Page 41 of 51Branch Technical Position RSLB-1, with periodic examinations and otherinspections of highly-affected internals locations. These examinations providereasonable assurance that the effects of age-related degradation mechanismswill be managed during the period of extended operation. The program includesexpanding periodic examinations and other inspections if the extent of thedegradation effects exceeds the expected levels.The MRP-227 guidance for selecting RVI components for inclusion in theinspection sample is based on a four-step ranking process. Through this process,the reactor internals were assigned to one of the following four groups: Primary,Expansion, Existing Programs, and No Additional Measures components.Definitions of each group are provided in GALL Chapter IX.B.The result of this four-step sample selection process is a set of Primary InternalsComponent locations for each of the three plant designs (Westinghouse,Combustion Engineering and Babcock & Wilcox) that are expected to show theleading indications of the degradation effects, with another set of ExpansionInternals Component locations that are specified to expand the sample shouldthe indications be more severe than anticipated. The degradation effects in athird set of internals locations are deemed to be adequately managed by ExistingPrograms. A fourth set of internals locations are deemed to require no additionalmeasures. As a result, the program typically identifies 5 to 15 percent of the RVIlocations as Primary Component locations for inspections, with another 7 to10 percent of the RVI locations to be inspected as Expansion Components, aswarranted by the evaluation of the inspection results. Another 5 to 15 percent ofthe internals locations are covered by Existing Programs, with the remainderrequiring no additional measures. This process thus uses appropriate componentfunctionality criteria, age-related degradation susceptibility criteria, and failureconsequence criteria to identify the components that will be inspected under theprogram in a manner that conforms to the sampling criteria for sampling-basedcondition monitoring programs in Section A.1.2.3.4 of NRC Branch PositionRLSB-1. Consequently, the sample selection process is adequate to assure thatthe intended function(s) of the PWR reactor internal components are maintainedduring the period of extended operation.No existing generic industry programs contain the specificity considered sufficientfor monitoring the aging effects addressed by the MRP-227 guidelines for B&Wplants. Therefore, no components for B&W plants were placed into the ExistingPrograms group.The program's use of visual examination methods in MRP-227 for detection ofrelevant conditions (and the absence of relevant conditions as a visualexamination acceptance criterion) is consistent with the ASME Code, Section Xlrules for visual examination. However, the program's adoption of the MRP-227
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AL-12-015Page 42 of 51guidance for visual examinations goes beyond the ASME Code, Section Xl visualexamination criteria because additional guidance is incorporated into MRP-227 toclarify how the particular visual examination methods will be used to detectrelevant conditions and describes in more detail how the visual techniques relateto the specific RVI components and how to detect their applicable age-relateddegradation effects.The technical basis for detecting relevant conditions using volumetric ultrasonictesting (UT) inspection techniques can be found in MRP-228, where the review ofexisting bolting UT examination technical justifications has demonstrated theindication detection capability of at least two vendors, and where vendortechnical justification is a requirement prior to any additional boltingexaminations. Specifically, the capability of program's UT volumetric methods todetect loss of integrity of PWR internals bolts, pins, and fasteners, such asbaffle-former bolting in B&W and Westinghouse units, has been welldemonstrated by operating experience. In addition, the program's adoption of theMRP-227 guidance and process incorporates the UT criteria in MRP-228, whichcalls for the technical justifications that are needed for volumetric examinationmethod demonstrations, required by the ASME Code,Section V.The program also includes future industry operating experience as incorporatedin periodic revisions to MRP-227. The program thus provides reasonableassurance for the long-term integrity and safe operation of reactor internals in allcommercial operating U.S. PWR nuclear power plants.Age-related degradation in the reactor internals is managed through anintegrated program. Specific features of the integrated program are listed in thefollowing ten program elements. Degradation due to changes in materialproperties (e.g., loss of fracture toughness) was considered in the determinationof inspection recommendations and is managed by the requirement to useappropriately degraded properties in the evaluation of identified defects. Theintegrated program is implemented by the applicant through an inspection plan.The Davis-Besse PWR Reactor Vessel Internals Program will address allplant-specific action items applicable to Davis-Besse that are established inSection 4.2 of the safety evaluation for MRP-227. In addition, a plant-specificinspection plan for ensuring the implementation of MRP-227 program guidelinesT,as amended by tho safoy oval.ati,,n Afor AdRip 227-, and Davis-Besse's responsesto the plant-specific action items, as identified in Section 4.2 of the safetyevaluationI ,Ar- MR2.2-J, will be submitted for NRC review and approval.
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AL-12-015Page 43 of 51NUREG-1801 ConsistencyThe PWR Reactor Vessel Internals Program is a new Davis-Besse program thatwill be consistent with the 10 elements of an effective aging managementprogram as described in NUREG-1801, Rev. 2,Section XI.M16A, "PWR VesselInternals." The results of an evaluation for each element are provided below.Exceptions to NUREG-1801None.EnhancementsNone.Aging Management Program ElementsThe results of an evaluation of each program element are provided below.ScopeThe scope of the program includes all RVI components at Davis-Besse, whichis built to a B&W NSSS design. The scope of the program applies themethodology and guidance in the most recently NRC-endorsed version ofMRP-227, which provides augmented inspection and flaw evaluationmethodology for assuring the functional integrity of safety-related internals incommercial operating U.S. PWR nuclear power plants designed by B&W, CE,and Westinghouse. The scope of components considered for inspectionunder MRP-227 guidance includes core support structures (typically denotedas Examination Category B-N-3 by the ASME Code,Section XI), those RVIcomponents that serve an intended license renewal safety function pursuantto criteria in 10 CFR 54.4(a)(1), and other RVI components whose failurecould prevent satisfactory accomplishment of any of the functions identified in10 CFR 54.4(a)(1)(i), (ii), or (iii). The scope of the program does not includeconsumable items, such as fuel assemblies, reactivity control assemblies,and nuclear instrumentation, because these components are not typicallywithin the scope of the components that are required to be subject to an agingmanagement review (AMR), as defined by the criteria set in10 CFR 54.21(a)(1). The scope of the program also does not include weldedattachments to the internal surface of the reactor vessel because thesecomponents are considered to be ASME Code Class 1 appurtenances to thereactor vessel and are adequately managed in accordance with an applicant'saging management program (AMP) that corresponds to GALL AMP XI.M1,"ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC,and IWD."
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AL-12-015Page 44 of 51In addition, the scope of the program includes management of thetime-limited aging analysis (TLAA) identified in LRA Section 4.2.7 forreduction in fracture toughness of the reactor vessel internals. This TLAA willbe managed in accordance with the implementation of the MRP-227guidelines, as amendod by thoeMRP 227 Safoty ovaluation, including allactivities associated with Davis-Besse's responses to plant-specific actionitems identified in the Section 4.2 of the safety evaluation.The scope of the program includes the response bases to applicable licenserenewal applicant action items (LRAAIs) on the MRP-227 methodology, andany additional programs, actions, or activities that are discussed in theseLRAAI responses and credited for aging management of the applicant's RVIcomponents. The LRAAIs are identified in the staffs safety evaluation onMRP-227 and include applicable action items on meeting those assumptionsthat formed the basis of the MRP's augmented inspection and flaw evaluationmethodology (as discussed in Section 2.4 of MRP-227), and NSSSvendor-specific or plant-specific LRAAIs as well. Davis-Besse's responses tothe plant-specific action items, as identified in Section 4.2 of the safetyevaluation for MRP-227, will be submitted for NRC review and approval.The guidance in Section 2.4 of MRP-227 specifies applicability limitations tobase-loaded plants and the fuel loading management assumptions uponwhich the functionality analyses were based. General assumptions used inthe analysis include:1) 30 years of operation with high leakage core loading patterns (freshfuel assemblies loaded in peripheral locations) followed byimplementation of a low-leakage fuel management strategy for theremaining 30 years of operation;2) base load operation, i.e., typically operates at fixed power levels anddoes not usually vary power on a calendar or load demand schedule;and3) no design changes beyond those identified in general industryguidance or recommended by the original vendors.Davis-Besse had approximately 13 years of operation with fresh fuelassemblies at peripheral locations. Cycle 15 has implemented a new failureresistant fuel design in high vulnerability locations. The core design forDavis-Besse is within the assumption of MRP-227. Davis-Besse is a baseload plant and has incorporated no design changes beyond those identified ingeneral industry guidance or recommended by the original vendors.
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AL-12-015Page 45 of 51* Preventive ActionsThe guidance in MRP-227 relies on PWR water chemistry control to preventor mitigate aging effects that can be induced by corrosive aging mechanisms(e.g., loss of material induced by general, pitting corrosion, crevice corrosion,or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]).Reactor coolant water chemistry is monitored and maintained in accordancewith the PWR Water Chemistry Program. The PWR Water ChemistryProgram is an existing Davis-Besse program that is consistent with the10 elements of an effective aging management program as described inNUREG-1801,Section XI.M2, "Water Chemistry."* Parameters Monitored or InspectedThe program manages the following age-related degradation effects andmechanisms that are applicable in general to the RVI components at thefacility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclicalloading; (b) loss of material induced by wear; (C) loss of fracture toughnessinduced by either thermal aging or neutron irradiation embrittlement;(d) changes in dimension due to void swelling and irradiation growth,distortion, or deflection; and (e) loss of preload caused by thermal andirradiation-enhanced stress relaxation or creep. For the management ofcracking, the program monitors for evidence of surface breaking lineardiscontinuities if a visual inspection technique is used as the non-destructionexamination (NDE) method, or for relevant flaw presentation signals if avolumetric UT method is used as the NDE method. For the management ofloss of material, the program monitors for gross or abnormal surfaceconditions that may be indicative of loss of material occurring in thecomponents. For the management of loss of preload, the program monitorsfor gross surface conditions that may be indicative of loosening in applicablebolted, fastened, keyed, or pinned connections. The program does notdirectly monitor for loss of fracture toughness that is induced by thermal agingor neutron irradiation embrittlement, or by void swelling and irradiation growth;instead, the impact of loss of fracture toughness on component integrity isindirectly managed by using visual or volumetric examination techniques tomonitor for cracking in the components and by applying applicable reducedfracture toughness properties in the flaw evaluations if cracking is detected inthe components and is extensive enough to warrant a supplemental flawgrowth or flaw tolerance evaluation under the MRP-227 guidance or ASMECode, Section Xl requirements. The program uses physical measurements tomonitor for any dimensional changes due to void swelling, irradiation growth,distortion, or deflection.Specifically, the program implements the parameters monitored/inspectedcriteria for B&W designed Primary Components in Table 4-1 of MRP-227.
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AL-12-015Page 46 of 51Additionally, the program implements the parameters monitored/inspectedcriteria for B&W designed Expansion Components in Table 4-4 of MRP-227.No existing generic industry programs contain the specificity consideredsufficient for monitoring the aging effects addressed by the MRP-227guidelines for B&W plants. Therefore, no components for B&W plants wereplaced into the Existing Programs group. No inspections, except for thosespecified in ASME Code, Section Xl, are required for components that areidentified as requiring "No Additional Measures," in accordance with theanalyses reported in MRP-227. As part of the Davis-Besse InserviceInspection Program, a visual VT-3 examination of the reactor vesselremovable core support structure is conducted once per Inservice Inspectioninterval in accordance with ASME Section Xl, Table IWB-2500-1, ExaminationCategory B-N-3.MRP-227 I&E guidelines require a visual (VT-3) examination of the coresupport shield (CSS) vent valve retaining rings and disc ahaft for every10year Inservice Inspection Interval. In addition, Davis-Besse TechnicalSpecification 5.5.4 requires testing of the CSS vent valves every 24 months toverify by visual inspection that the valve body and valve disc exhibit noabnormal degradation, verify the valve is not stuck in an open position, andverify by manual actuation that the valve is fully open when a force of<400 lbs is applied vertically upward. The technical specification inspectionwill continue to be performed at the prescribed frequency of 24 months. TheMRP-227 required visual (VT-3) examination will also be performed at theprescribed frequency of every 10 year Inservice Inspection Interval.Detection of Aging EffectsThe detection of aging effects is covered in two places: (a) the guidance inSection 4 of MRP-227 provides an introductory discussion and justification ofthe examination methods selected for detecting the aging effects of interest;and (b) standards for examination methods, procedures, and personnel areprovided in a companion document, MRP-228. In all cases, well-establishedmethods were selected. These methods include volumetric UT examinationmethods for detecting flaws in bolting, physical measurements for detectingchanges in dimension, and various visual (VT-3, VT-1, and EVT-1)examinations for detecting effects ranging from general conditions todetection and sizing of surface-breaking discontinuities. Surface examinationsmay also be used as an alternative to visual examinations for detection andsizing of surface-breaking discontinuities.Cracking caused by SCC, IASCC, and fatigue is monitored/inspected byeither VT-1 or EVT-1 examination (for internals other than bolting) or byvolumetric UT examination (bolting). The VT-3 visual methods may be applied
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AL-12-015Page 47 of 51for the detection of cracking only when the flaw tolerance of the component oraffected assembly, as evaluated for reduced fracture toughness properties, isknown and has been shown to be tolerant of easily detected large flaws, evenunder reduced fracture toughness conditions. In addition, VT-3 examinationsare used to monitor/inspect for loss of material induced by wear and forgeneral aging conditions, such as gross distortion caused by void swellingand irradiation growth or by gross effects of loss of preload caused by thermaland irradiation-enhanced stress relaxation and creep.In addition, the program adopts the recommended guidance in MRP-227 fordefining the Expansion criteria that need to be applied to inspections ofPrimary Components and Existing Requirement Components and forexpanding the examinations to include additional Expansion Components. Asa result, inspections performed on the RVI components are performedconsistent with the inspection frequency and sampling bases for PrimaryComponents and Expansion Components in MRP-227, which have beendemonstrated to be in conformance with the inspection criteria, samplingbasis criteria, and sample Expansion criteria in Section A.1.2.3.4 of NRCBranch Position RLSB-1.Specifically, the program implements the parameters monitored/inspectedcriteria and bases for inspecting the relevant parameter conditions for B&Wdesigned Primary Components in Table 4-1 of MRP-227 and for B&Wdesigned Expansion Components in Table 4-4 of MRP-227.As provided in Soctien 4.4.3 of the MRP 227 sapey evalan, theasrme'disrbutse t sho l efotaging bolte (aiso known as the flow destebutsr boes) inRA 144 designed plants Wares dded to the relmain inspegtien d ategnsio Thesafety evaluatien pro-vides-Q th-at the ex-aminationa method- shaoll bhe volume~tricexaminatid n (UT), the examnnatien coverage for these aemponents shaolponfo to the acrieria as desrogbed in Section 3.3.41 of the saof evalRat-n,and the Fe examination frequency shall be on a 40 year- int-ernal sAimilar to1olther 'PrhMary" inspection caegory componsents. For BR&W designed plants,,Pno other a;;dditinal_ components were- added- to the "Priar-y" insectioncategor-y and no add-itional ~omponents Iere a~dded to the- "Expansion"insection categeor/In addition, in some cases (as defined in MVRP-227), physical measurementsare used as supplemental techniques to manage for the gross effects of wear,loss of preload due to stress relaxation, or for changes in dimension due tovoid swelling, deflection or distortion. The physical measurements methodsapplied in accordance with this program includes Section 4.3.1 of MRP-227that describes the physical measurements needed for the B&W internals coreclamping items. In addition, Table 4-1 provides the required examination
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AL-12-015Page 48 of 51method and examination coverage and Table 5-1 provides the acceptancecriteria for the physical measurements.Monitoring and TrendingThe program requires that all inspections shall be documented for futurereview; defects shall be documented in accordance with the Davis-Bessecorrective action program.In addition, the program requires that a summary report of all inspections andmonitoring, items requiring evaluation, and new repairs shall be submitted tothe MRP Program Manager within 120 days of the completion of an outageduring which PWR internals within the scope of MRP-227 are examined.Section 6 of MRP-227 will not be used by FENOC for evaluating examinationresults that do not meet the acceptance criteria identified in Section 5 ofMRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as theframework to develop those generic and plant-specific evaluations triggeredby findings in the RVI examinations. As provided in the safety evaluation forMRP-227, Rev-0-, the NRC staff is currently reviewing WCAP-17096-NP,Revision 2." Acceptance CriteriaSection 5 of MRP-227 provides specific examination acceptance criteria forthe Primary and Expansion Component examinations. For componentsaddressed by examinations referenced to ASME Code, Section Xl, theIWB-3500 acceptance criteria apply. For other components covered byExisting Programs, the examination acceptance criteria are described withinthe Existing Program reference document.The guidance in MRP-227 contains three types of examinationacceptance criteria:For visual examination (and surface examination as an alternative tovisual examination), the examination acceptance criterion is theabsence of any of the specific, descriptive relevant conditions; inaddition, there are requirements to record and disposition surfacebreaking indications that are detected and sized for length byVT-1/EVT-1 examinations;For volumetric examination, the examination acceptance criterion isthe capability for reliable detection of indications in bolting, asdemonstrated in the examination Technical Justification; in addition,there are requirements for system-level assessment of bolted or
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AL-12-015Page 49 of 51pinned assemblies with unacceptable volumetric (UT) examinationindications that exceed specified limits; and,For physical measurements, the examination acceptance criterion forthe acceptable tolerance in the measured differential height from thetop of the plenum rib pads to the vessel seating surface in B&W plantsare given in Table 5-1 of MRP-227.Section 6 of MRP-227 will not be used by FENOC for evaluating examinationresults that do not meet the acceptance criteria identified in Section 5 ofMRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as theframework to develop those generic and plant-specific evaluations triggeredby findings in the RVI examinations. As provided in the safety evaluation forMRP-227, Rev- OL, the NRC staff is currently reviewing WCAP-17096-NP,Revision 2." 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 programs and activities that arecredited with aging management during the period of extended operation andis discussed in Section B .1.3.* Operating ExperienceRelatively few incidents of PWR internals aging degradation have beenreported in operating U.S. commercial PWR plants. However, a considerableamount of PWR internals aging degradation has been observed in EuropeanPWRs, with emphasis on cracking of baffle-former bolting. For this reason,the U.S. PWR owners and operators began a program a decade ago toinspect the baffle-former bolting in order to determine whether similarproblems might be expected in U.S. plants. A benefit of this decision was theexperience gained with the UT examination techniques used in theinspections. In addition, the industry began substantial laboratory testingprojects in order to gather the materials data necessary to support futureinspections and evaluations. Another item with existing or suspected material
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AL-12-015Page 50 of 51degradation concerns that has been identified for PWR components iscracking in some high-strength bolting. This condition has been correctedprimarily through bolt replacement with less susceptible material andimproved control of pre-load.Stress corrosion cracking (SCC) has occurred in Alloy A-286 internals boltingin B&W units, this included Davis-Besse. The Alloy A-286 bolt failures in B&WPWR internals were subjected to a comprehensive failure analysis that isdocumented in BAW-1843PA, "The B&W Owners Group Evaluation ofInternal Bolting Concerns in 177FA Plants," dated January 1986.BAW-1843PA was reviewed and approved by the NRC. This failure analysisaddressed probable cause of the cracking, assessment of likelihood andconsequences of joint failure, and replacement bolt design. Therecommended replacement bolts were Alloy X-750 HTH bolts that are lesssusceptible to SCC and have overall excellent material properties.Davis-Besse has replaced the majority of the Alloy A-286 bolts for the reactorvessel internals (upper core barrel, lower core barrel, lower thermal shieldand surveillance specimen holder tubes) with Alloy X-750 HTH bolts. Tosatisfy a needed action under NEI 03-08 protocol, Davis-Besse performed UTexaminations of 100% of all upper core barrel bolts during the cycle 16refueling outage. This inspection did not identify any unacceptableindications.As part of the Inservice Inspection Program, a visual (VT-3) examination ofthe reactor vessel removable core support structure is conducted once perInservice Inspection interval in accordance with ASME Section Xl, TableIWB-2500-1, Examination Category B-N-3. These inspections have notidentified any unacceptable indications.FENOC participates in the industry programs for investigating and managingaging effects on reactor vessel internals. Through its participation in EPRIMRP activities, FENOC will continue to benefit from the reporting of reactorvessel internals inspection information, and will share its own internalsinspection results with the industry, as appropriate.ConclusionThe PWR Reactor Vessel Internals Program provides reasonable assurance thatcracking, including stress corrosion cracking (SCC), which also encompassesprimary water stress corrosion cracking (PWSCC), irradiation-assisted stresscorrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; loss ofmaterial induced by wear; loss of fracture toughness due to either thermal agingor neutron irradiation embrittlement; and loss of preload due to thermal and
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AL-12-015Page 51 of 51irradiation-enhanced stress relaxation or creep of the subject reactor vesselinternals components will be adequately managed so that intended functions ofcomponents within the scope of license renewal are maintained consistent withthe current licensing basis for the period of extended operation.
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BDavis-Besse Nuclear Power Station, Unit No. I (DBNPS)Letter L-12-015New and Revised DBNPS License Renewal Application Boundarv Drawinqs14 pages followThe following License Renewal Application Boundary Drawingsare new and are enclosed:LR Drawing LR-MO10D Sheet 2LR Drawing LR-M036C Sheet 2LR Drawing LR-M039B Sheet 2Revision 0Revision 0Revision 0The following License Renewal Application Boundary Drawingsare revised and are enclosed:LR Drawing LR-MO10D Sheet ILR Drawing LR-MO12E,LR Drawing LR-M020BLR Drawing LR-M036C Sheet ILR Drawing LR-M037CLR Drawing LR-M037DLR Drawing LR-M037ELR Drawing LR-M037FLR Drawing LR-M039ALR Drawing LR-M039B Sheet 1LR Drawing LR-M042BRevision 3Revision 2Revision 3Revision 3Revision 2Revision 4Revision 3Revision 2Revision 4Revision 3Revision 3 The 14 Drawingsspecifically. referenced inEnclosure B have beenprocessed into ADAMSThese drawings can beaccessed by the NRC staffwithin the ADAMS packageor by performing a searchon the Document/ReportNumberr' D01=D14