IR 05000219/2012007

IR 05000219-12-007, on 01/23/2012-02/10/2012, Exelon Nuclear; Oyster Creek Generating Station, Engineering Specialist Plant Modifications Inspection

ML120790052
Person / Time
Site:	Oyster Creek
Issue date:	03/16/2012
From:	Doerflein L T Engineering Region 1 Branch 2
To:	Pacilio M J Exelon Generation Co, Exelon Nuclear
References
IR-12-007
Download: ML120790052 (22)

Text

ffi l.larch 76, 2AL2Mr. MichaelJ. PacilioSenior Vice President, Exelon Generation Company, LLCPresident and Chief Nuclear Officer, Exelon Nuclear4300 Winfield RoadWarenville. lL 60555

SUBJECT: OYSTER CREEK GENERATING STATION _ NRC EVALUATION OFCHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANTMODI FI CATI ONS TEAM I N SPECTI ON REPORT O5OOO2 1 91201 2OO7

Dear Mr. Pacilio:

On February 10, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed aninspection at the Oyster Creek Generating Station. The enclosed inspection report documentsthe inspection results, which were discussed on February 10,2012, with Mr. M. Massaro,Site Vice President, and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.ln conducting the inspection, the team reviewed selected procedures, calculations and records,observed activities, and interviewed station personnel.Based on the results of this inspection, no findings were identified.ln accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in theNRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, Agencywide Documents Access and Management System (ADAMS).ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (thePublic Electronic Reading Room).

Sincerely,Engineering Branch 2Division of Reactor SafetyDocket No. 50-219License No. DPR-16

Enclosure:

f nspection Report 050002 1 912012007M

Attachment:

Supplemental Informationcc Mencl: Distribution via ListServ

SUMMARY OF FINDINGS

lR 0500021912012007;0112312012 - 0211012012; Oyster Creek Generating Station; EngineeringSpecialist Plant Modifications Inspection.This report covers a two week on-site inspection period of the evaluations of changes, tests, orexperiments and permanent plant modifications. The inspection was conducted by three regionbased engineering inspectors. The NRC's program for overseeing the safe operation ofcommercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"Revision 4, dated December 2006.No findings were identified.Enclosure

REPORT DETAILS

1. REACTORSAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R17 Evaluations of Chanses. Tests, or Experiments and Permanent Plant Modifications(tP 71111.17).1 Evaluations of Chanqes. Tests. or Experiments (25 samples)a. Inspection ScopeThe team reviewed five safety evaluations to determine whether the changes to thefacility or procedures, as described in the Updated Final Safety Analysis Report(UFSAR), had been reviewed and documented in accordance with 10 CFR 50.59requirements. In addition, the team evaluated whether Exelon had been required toobtain NRC approval prior to implementing the changes. The team interviewed plantstaff and reviewed supporting information including calculations, analyses, designchange documentation, procedures, the UFSAR, the Technical Specifications (TS), andplant drawings to assess the adequacy of the safety evaluations. The team comparedthe safety evaluations and supporting documents to the guidance and methods providedin Nuclear Energy Institute (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," asendorsed by NRC Regulatory Guide 1.187 , "Guidance for lmplementation of10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of thesafety evaluations.The team also reviewed a sample of twenty 10 CFR 50.59 screenings for which Exelonhad concluded that no safety evaluation was required. These reviews were performedto assess whether Exelon's threshold for performing safety evaluations was consistentwith 10 CFR 50.59. The sample included design changes, calculations, and procedurechanges.The team reviewed the safety evaluations that Exelon had performed and approvedduring the time period covered by this inspection (i.e., since the last modificationsinspection) not previously reviewed by NRC inspectors. The 10 CFR 50.59 screeningswere selected based on the safety significance, risk significance, and complexity of thechange to the facility.In addition, the team compared Exelon's administrative procedures used to controlthescreening, preparation, review, and approval of safety evaluations to the guidance in NEI96-07 to determine whether those procedures adequately implemented the requirementsof 10 CFR 50.59. The safety evaluations and screenings reviewed by the team arelisted in the Attachment.b. FindinosNo findings were identified.Enclosure

.2.33 length spool pieces would be needed for the modification, requiring additional pipeflanges, resulting in increased loads on existing pipe supports and increased pipestresses.The team reviewed the modification to verify that the design and licensing bases andstructural integrity of the ESW piping and supports had not been degraded by themodification. The team interviewed design engineers, and reviewed evaluations, pipestress calculations, surveillance and PMT results, and associated maintenance workorders to verify that the ESW piping spool replacements were appropriately implementedand that the ESW piping was maintained in accordance with design assumptions. Theteam also performed several walkdowns of the accessible portions of the modification toensure that the system configuration was in accordance with design instructions and thatESW piping integrity was maintained. The 10 CFR 50.59 screening determinationassociated with this modification was reviewed as described in section 1R17 .1 of thisreport. The team also reviewed corrective action lRs to determine if there were reliabilityor performance issues that may have resulted from the modification. The documentsreviewed are listed in the Attachment.FindinqsNo findings were identified.lntake Trash Rake Upqrade ModificationInspection ScopeThe team reviewed modification 06-00819 that replaced the existing rail mounted andconveyor debris removal system at the intake structure with a new upgraded trash rakingsystem. The new system consists of bar racks, raking head and trolley, overheadmonorail and associated support columns, dedicated trash dumpster, and an automatedcontrol system with remote emergency stop capability. The previous rail mountedsystem proved difficult to operate and ineffective for collection and disposal of intakedebris especially during periods of high debris accumulation. Exelon upgraded theintake debris removal system with the newer, automated design to improve the system'seffectiveness and reliability.The team reviewed the modification to verify that the design and licensing bases andperformance capability of the intake structure and its supported systems had not beendegraded by the modification. Specifically, the team reviewed calculations, technicalevaluations, and operating procedures to verify that the overhead monorail andassociated support columns would not adversely impact important to safety structures,systems, and components (SSC) at the intake during normal operation or under designbasis conditions. The team reviewed the associated work order instructions anddocumentation to verify that maintenance personnel implemented the modification asdesigned. The team reviewed the associated PMT results, interviewed plant operators,and directly observed debris removal activities at the intake to verify proper operation ofthe upgraded system. The team also performed several walkdowns of the upgradedEnclosure

4rake system and intake area SSCs to ensure that maintenance personnel installed themodification as designed, and to independently assess Exelon's configuration controland the material condition of the intake area. In addition, the 10 CFR 50.59 screeningdetermination associated with this modification was reviewed as described in section1 R17.1 of this report. The team also reviewed corrective action lRs to determine if therewere reliability or performance issues that may have resulted from the modification. Thedocuments reviewed are listed in the Attachment.b. FindinqsNo findings were identified..2.4 Emeroencv Service Water Pump 'C' Discharoe Pipinq Modificationa. lnspection ScopeThe team reviewed modification 11-00035 that installed an uncoated flanged teedownstream of the 'C' ESW pump discharge. The piping tee was physically locatedbeneath the intake structure deck. The piping tee replacement was emergent worknecessitated due to an Exelon-identified through-wall leak in the existing piping tee.Exelon performed the replacement under a TS Limiting Condition for Operation that didnot allow sufficient time for Exelon to obtain an internally coated tee. Accordingly,engineering evaluated the acceptability of using the uncoated piping tee until Exelon'splanned replacement of all the ESW piping under the intake structure deck during theF all 2012 refueling outage.The team reviewed the modification to verify that the design and licensing bases andstructural integrity of the ESW piping had not been degraded by the modification. Theteam interviewed design engineers, and reviewed evaluations, non-destructiveexamination results, surveillance and PMT results, and associated maintenance workorders to verify that the ESW piping tee replacement was appropriately implemented,and that the ESW piping configuration supported continued operability throughDecember 2012. The team also performed several walkdowns of the accessibleportions of the modification to ensure that the system configuration was in accordancewith design instructions and that ESW piping integrity was maintained. The10 CFR 50.59 screening determination associated with this modification was reviewedas described in section 1R17.1 of this report. The team also reviewed corrective actionlRs to determine if there were reliability or performance issues that may have resultedfrom the modification. The documents reviewed are listed in the Attachment.b. FindinosNo findings were identified.Enclosure b.a.2.65Weak Link Analvsis for Ovster Creek Motor-Operated ValvesInspection ScopeThe team reviewed a modification 09-00889 that revised a weak link analysis (C-1302-900-E540-020) for several motor-operated valves (MOVs), including core spray systemMOVs. The weak link analysis was revised to address valve stem material changes aswell as a conversion to valve stems containing integral gages for MOV diagnostic testingfor several of the MOVs.The team reviewed the weak link analysis to verify that the design and licensing basesand performance capability of the MOVs not been degraded by the change. The teaminterviewed engineering staff and reviewed the revised analysis to confirm the impactedsystems would function in accordance with the design assumptions. The team alsoreviewed the corrective action lR database to determine if there were reliability orperformance issues that may have resulted from the modification. ln addition, the teamreviewed the associated equivalency change documentation that demonstrated that aformal 10 CFR 50.59 screen was not required. The documents reviewed are listed inthe Attachment.FindinqsNo findings were identified.Service Water Cross-Connect from Reactor Buildinq Closed Coolinq Water HeatExchanqer to Emerqencv Service WaterInspection ScopeThe team reviewed a modification 09-00433 that cross-connected the service water(SW) system reactor building closed cooling water (RBCCW heat exchanger dischargewith the existing ESWSW cross-connect piping. The intent of the modification was toallow the replacement of the SW system piping downstream of the RBCCW heatexchanger leading into the SW system discharge header and prevent entry into high riskplant configuration during the refueling outage. This configuration allowed sufficientcooling to the RBCCW system while the SW normaldischarge piping was replaced.The team reviewed the modification to verify that the design and licensing bases andstructural integrity of the associated piping had not been degraded by the modification.The team interviewed design engineers, and reviewed evaluations, examination results,and associated completed maintenance activities to verify that the modified pipingconfiguration supported continued functionality. The team performed walkdowns of themodification to ensure that the system configuration was in accordance with designinstructions. The team also reviewed corrective action lRs to determine if there werereliability or performance issues that may have resulted from the modification. InEnclosure 6addition, the 10 CFR 50.59 screening determination associated with this modificationwas reviewed as described in section 1R17.1 of this report. The documents reviewedare listed in the Attachment.b. FindinqsNo findings were identified..2.7 Hardened Vent Valve Open Position to Permit Ventinqa. Inspection ScopeThe team reviewed modification 08-00864, which involved a calculation revision, todetermine the smallest acceptable valve opening for the hardened vent system valvesthat would allow the required venting for both the drywell and torus. The stroke for thesubject butterfly valves, V-23-13, -14, -15, and -16, had been limited to 75 degrees toallow them to close in the required stroke time, and which also considered valvestructural limitations. However, there was no tolerance associated with setting up thevalve stroke. This calculation determined the minimum valve position, to be used forsetup tolerance, to ensure that the vent capability would be satisfied. The valves servetwo purposes: 1) to close to provide containment isolation, and 2) to provide a hardenedvent function. The valves are now required to have a 70 degrees - 75 degrees band forthe open position limit.The team reviewed the calculation to verify that the design and licensing bases andperformance capability of the containment isolation and hardened vent functions of thesubject valves had not been degraded by the modification. Specifically, the teamverified that design specifications remained valid for postulated scenarios. The teaminterviewed engineers, and reviewed evaluations and completed surveillance and in-service test results to verify that the open position tolerances were appropriatelyimplemented. The 10 CFR 50.59 screening determination associated with thismodification was reviewed as described in section 1R17.1 of this report. Finally, theteam walked down the isolation valves with the system engineer to assess the materialcondition of the valves. The documents reviewed are listed in the Attachment.b. FindinqsNo findings were identified..2.8 Calculation on Combustion Turbine Tank Oil Levela. lnspection ScopeThe team reviewed modification 10-00175 that revised calculation C-1302-743-E310-006 related to the required minimum fuel oil level for the Forked River CombustionTurbine (FRCT) fuel tank. The fuel reserve is required by contract with an outsideorganization to assure that the FRCTs can provide the power to Oyster Creek in theEnclosure 7event of a station blackout. The existing minimum level in the fueltank was 14 feet, butthe revision changed the minimum level to 8 feet.The team reviewed the calculation to verify that the design and licensing bases andperformance capability of the FRCT functions had not been degraded by themodification. Specifically, the team verified that design specifications remained valid forthe postulated station blackout scenario. The team interviewed design engineers andreviewed the existing contract to ensure scenario assumptions remained valid. The10 CFR 50.59 screening determination associated with this modification was reviewedas described in section 1R17.1 of this report. Finally, the team walked down the FRCTS,including the fueltank and associated fuel supply system with the responsible engineerto assess the material condition of the FRCT system. The documents reviewed arelisted in the Attachment.b. FindinqsNo findings were identified..2.9 Emerqencv Diesel Generator Batterv Voltaqe for Control Circuitsa. lnspection ScopeThe team reviewed modification 07-00552, which created a calculation to determine theminimum voltage available for both emergency diesel generator (EDG) breaker closingcoils and associated control devices. The EDGs provide vital power to emergencybuses during a loss-of-offsite power. Each EDG has a dedicated 120Vdc battery toprovide starting power. The modification was performed because Exelon received afinding (NCV 05000219/2007006-01) in 2007 during a Component Design BasisInspection performed by the NRC at Oyster Creek. The finding was related to anexisting EDG battery sizing calculation that did not address the available DC voltage tothe EDG breaker closing coils or associated control devices. The modification did notinclude any physical plant changes to the facility.The review was performed to verify that the design and licensing bases of the facility hadnot been degraded by the results of the new calculation. The team reviewed thecalculation and technical evaluations to assess whether the modification was consistentwith design assumptions. Power requirements were reviewed to verify that the EDGbreaker closing coils and associated control circuits met the manufacturer'sspecifications. Supporting electrical calculations and analyses for the EDG batterysizing requirements were reviewed to ensure design limits were not exceeded. Theteam performed a walkdown of the EDG battery compartments to identify any abnormalconditions while in service. The team also conducted interviews with engineering staff todetermine if the affected SSCs would function in accordance with the designassumptions. Finally, the 10 CFR 50.59 screening determination associated with thismodification was reviewed as described in section 1 R17.1 of this report. The documentsreviewed are listed in the Attachment.Enclosure 8b. FindinqsNo findings were identified..2.10 Revise Ovster Creek Short Circuit Studva. Inspection ScopeThe team reviewed modification 07-00744 that updated the Oyster Creek Short CircuitStudy (C-1302-700-5350-012). The modification included converting the existing ShortCircuit Study modeled in electricaltransient analysis software (DAPPER) to newelectrical transient analysis software (ETAP). The new electrical distribution model alsoincluded corrections to reflect the as-built plant configuration. An additional scenario,where the main generator provides power to the auxiliary transformer with one of theEDGs connected in parallelto an emergency bus for EDG testing, was also included.The modification did not include any physical plant changes to the facility.The review was performed to verify that the design bases and licensing bases of thefacility had not been degraded by the short circuit study results. The results of therevised short circuit study showed that the interrupting and momentary fault currentswere within the circuit breaker ratings of allthe 4160V Switchgear, 480V unitsubstations, and 480V motor control centers. There was also an improvement incalculated margin for the worst case scenario of a small break loss-of-coolant accident.Design assumptions were reviewed to evaluate whether they were technicallyappropriate and consistent with the UFSAR. The team also conducted interviews withengineering staff to determine if the affected SSCs would function in accordance with thedesign assumptions. Finally, the 10 CFR 50.59 screening determination associated withthis modification was reviewed as described in section 1R17.1 of this report. Thedocuments reviewed are listed in the Attachment.b. FindinssNo findings were identified..2.11 Plant Process Computer lntelliqent Remote Control Unit Replacementa. Inspection ScopeThe team reviewed modification 08-00527 that upgraded the inpuUoutput components ofthe plant process computer. Two intelligent remote control units (IRCU) and sixexpansion chassis were replaced with eight universal chassis. The existing inpuUoutputcards were transferred from the existing IRCU and expansion chassis to the newuniversal chassis. The modification was performed because the existing system used acustom interface that on rare occasion locked up the system and changed the scannedvalues to static values. The universal chassis interface with data from the followingsystems: Rod Worth Minimizer; 3D Monicore; Safety Parameter Display System;Enclosure IEmergency Response Data System; and Radioactive Gaseous Effluent MonitoringSystem.The review was performed to verify that the design and licensing bases andperformance capability of the installed universal chassis had not been degraded by themodification. Power requirements were reviewed to verify that the installed universalchassis met the manufacturer's specifications. Replacement components and materialswere reviewed to ensure that the modification conformed to the design specifications.The team also verified that selected drawings and calculations were properly updatedbased on the installed universal chassis. The team reviewed the PMT to verify properoperation of the installed universal chassis. The team reviewed lRs associated with theuniversal chassis to verify that deficiencies were appropriately identified and corrected.The team also conducted interviews with engineering staff to verify that the affectedSSCs functioned in accordance with the design assumptions, and to verify themodification corrected the previously identified problem. The 10 CFR 50.59 screeningdetermination associated with this modification was reviewed as described in section1R17

.1 of this report. The documents reviewed are listed in the Attachment.b. FindinqsNo findings were identified..2.12 Replace Recorder UR-622-24B for Post-Accident Monitorinq Reactor Pressure andLevela. Inspection ScopeThe team reviewed modification 09-00590 that replaced an existing Yokogawa analogstrip chart recorder with a new Yokogawa digital paperless recorder. The new digitalrecorder displays the same plant inputs as the existing analog recorder. The followingdata inputs are displayed on the new recorder: Reactor Pressure; Reactor Level-Yarway; and Reactor Level- Fuel Zone. The modification was performed because theexisting analog recorder was obsolete and spare parts were difficult to obtain. Themodification included installation of the new recorder in the same location as the existingrecorder and no additionalwiring was required.The review was performed to verify that the design and licensing bases andperformance capability of the new digital recorder had not been degraded by themodification. The team reviewed calculations and technical evaluations to assesswhether the modification was consistent with design assumptions. Power requirementswere reviewed to verify that the new digital recorder met the manufacturer'sspecifications. The replacement component was reviewed to verify that it wasseismically qualified. The team also verified that selected drawings, calculations,instrument calibration sheets, and procedures were properly updated based on theinstallation of the digital recorder. The team reviewed the PMT to verify proper operationof the digital recorder. The team reviewed lRs associated with the digital recorder toverify that deficiencies were appropriately identified and corrected. In addition, the teamEnclosure

4.10reviewed the associated equivalency change documentation that demonstrated that aformal 10 CFR 50.59 screen was not required. Finally, the team conducted interviewswith engineering staff to determine if the affected SSCs would function in accordancewith the design assumptions. The documents reviewed are listed in the Attachment.FindinssNo findings were identified.OTHER ACTIVITIESldentification and Resolution of Problems (lP 71152)Inspection ScopeThe team reviewed a sample of lRs associated with 10 CFR 50.59 and plantmodification issues to determine whether Exelon was appropriately identifying,characterizing, and correcting problems associated with these areas, and whether theplanned and/or completed corrective actions were appropriate. In addition, the teamreviewed lRs written on issues identified during the inspection to verify adequateproblem identification and incorporation of the problem into the corrective action system.The lRs reviewed are listed in the Attachment.FindinqsNo findings were identified.Meetinqs, includino ExitThe team presented the inspection results to Mr. M. Massaro, Site Vice President, andother members of Exelon's staff at an exit meeting on February 10,2012. The teamreturned the proprietary information reviewed during the inspection and verified that thisreport does not contain proprietary information.4c.4240A6b.Enclosure A-1ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTAGTExelon Personnel

A. Aganral, Design Engineer

M. Benitez, Design Engineer

J. Chrisley, Regulatory Assurance Specialist

P. De, Design Engineer

P. Desai, Design Engineer

J. Flores, Design Engineer

G. Malone, Director, Engineering

M. Massaro, Site Vice-President

K. Mayle, Design Engineer

P. Procacci, Design Engineer

H. Ray, Senior Manager, Design Engineering

T. Ruggiero, System Manager

S. Schwartz, System Manager

J. Tabone, MOV/AOV Program OwnerLIST OF ITEMS OPENED, CLOSED AND DISCUSSEDNoneLIST OF DOCUMENTS REVIEWED10 CFR 50.59 EvaluationsOC-2009-E-0001, SW Cross-Connect Downstream of RBCCW Heat Exchangers, Rev. 1OC-2010-E-0001, Application of TRACG04P Version 4.2.60.3 for Stability Analysis, Rev. 0OC-2010-E-0002, Transition to GNF-2 Fuel - lmpact on EAB, LPZ and CR Dose, Rev. 0OC-2011-E-0001, MSO Bus 1D Appendix R Permissive Switch lnstallation, Rev. 0OC-2011-E-0002, Core Spray System MSO Modifications - MSO Scenarios 5k & 10b, Rev. 010 CFR 50.59 Screened-out EvaluationsOC -2010-5-0046, Bank 5-6 Voltage Regulator Controller Modification, Rev. 0OC -2010-S-0108, Component Design Basis Inspection Calculation Revision, Rev. 0OC -2011-5-0004, EDG-2 Breaker Logic Modification for Appendix R Fire, Rev. 0OC -2011-5-0016, RPS Sub-Channel '1A'Alternate MSIV Closure Trip Signal, Rev. 0OC -2011-S-0017, lmplement Level 2-3Data Diodes, Rev. 0OC -2011-5-0048, Replacement of EDG-1 Speed Switch, Rev. 0OC -2011-S-0108, Core Spray Pump Trip Logic Mod - MSO Scenario 79, Rev. 0Attachment

A-2OC-2009-S-0093, 205.95.0 - Reactor Flood-up / Drain-down, Rev. 0OC-2009-S-0170,310 - Containment Spray System Operation, Rev. 0OC-2009-S-0179,681.4.005 - Substation Tour Sheet, Rev. 0OC-2009-S-0181, ECR 09-00678 - Replace Recorders LR-37/PR-53 and AR-1, Rev. 0OC-2009-S-0184,316.1 - Condensate Transfer System, Rev. 0OC-2010-S-0026, 301.2 - Reactor Recirculation System, Rev. 0OC-2010-5-0027,654.2.002 - Installation of Space Heater in Shutdown Cooling Room, Rev. 0OC-2010-5-0048, ECR 10-00175 FRCT Fuel Oil System Evaluation for SBO, Rev. 0OC-2010-5-009355, ECR 10-00108 - TBCCW Heat Exchanger Vent Bypass Line, Rev. 0OC-2010-S-0190, Surveillance 665.5.00 Allow use of Hardened Vent for Venting, Rev. 1OC-2011-5-0060, ABN-31 High Winds, Rev. 0OC-2011-5-0070, ABN-32 Abnormal Intake Level, Rev. 0OC-201 1-S-01 14. V-14-33 Steam Leak Condenser ECR 1 1-00596. Rev. 0Modification Packaqes06-00819, lntake Trash Rake Upgrade Modification, Rev. 007-00552, EDG Battery Voltage for Control Circuits, Rev. 007-00744, Revise Oyster Creek Short Circuit Study, Rev. 0Q8-00527, Plant Process Computer Intelligent Remote Control Unit Replacement, Rev. 208-00864, Hardened Vent Valve Open Position to Permit Venting, Rev. 008-01040, ESW System Piping Spool Replacements, Rev. 009-00433, SW Cross-Connect from RBCCW Heat Exchanger to ESW, Rev. 309-00590, Replace Recorder UR-622-24B for Post-Accident Monitoring Reactor Pressure andLevel, Rev. 009-00708, 'A'Control Room HVAC System Modifications, Rev. 009-00889, Weak Link Analysis for'Oyster Creek MOVs, Rev. 010-00175, Calculation on Combustion Turbine Tank Oil Level, Rev. 011-00035, ESW Pump 'C' Discharge Piping Modification, Rev. 0Calculations. Analvsis. and Evaluations1164020-07, ESW Piping Leak Under Deck at lntake Equipment (Apparent Cause), Rev. 01209774-02, ESW Pipe ES0106 lnspection Technical Evaluation, dated 61151111209774-04, ESW Pipe ES0112 Inspection Technical Evaluation, dated 61151111209774-05, ESW Pipe ES0212 Inspection Technical Evaluation, dated 6115111943876-03, Intake and Dilution Trash Rake Short Hose Failures (Apparent Cause), Rev. 042204999-05, CR HVAC'A'Condenser Support Bracket Seismic Evaluation, dated 9l24l0g42204999-15, Provide Guidance/Acceptance forAs-Built Condenser Clamps, dated 1111210942287516,'B' lC Steam Inlet lsolation Valve V-14-33 Packing Leakage Monitoring Plan, Rev. 242296464-01, Leakage from 'A' Control Room HVAC Technical Evaluation, dated 1131112C-1302-168-E310-001, lntake Structure Trash Rake Foundation and Anchorages, Rev. 1C-1302-251-E310-037, Pipe Stress Analysis, Fuel Pool Cooling Line NN-1, Model 2,fromSkimmer Surge Tank to Pump Suction Nozzles NN01A/B and NN01C/D, Rev. 0C-1302-532-E540-037, Piping Analysis - ESW System from Containment Spray HX 3 & 4through TB and Out to Yard, Rev. 1aC-1302-532-E540-041, Piping Analysis - ESW System from Containment Spray HX1 &2through RB to TB Entrance, Rev. 2aAttachment

A-3C-1302-532-E540-045, Piping Analysis - ESW System from Containment Spray HX 1 & 2 Outletsto Yard, Rev. 2aC-1302-700-5350-004, Oyster Creek Electrical Model, Rev. 3C-1302-700-5350-012, Oyster Creek Short Circuit Study, Rev. 3C-1302-741-5350-009, Oyster Creek EDG Sizing Calculation, Rev. 1C-1302-743-E310-006, Forked River Fuel Oil Transfer System Performance Evaluation, Rev. 0C-1302-822-5360-036, Hardened Vent System Capability, Rev. 2C-1302-822-E310-083, EAB, LPZ, Control Room Doses due to LOCA, Rev. 0C-1302-822-E310-084, EAB, LPZ, CR Doses due to Main Steam Line Break Accident, Rev. 0C-1302-822-E310-085, EAB, LPZ, CR Doses due to Fuel Handling Accident, Rev. 0C-1302-822-E310-086, EAB, LPZ, CR Doses due to Control Rod Drop Accident, Rev. 0C-1302-900-E540-020, Weak Link Analysis Calculations for Oyster Creek MOVs, Rev. 1ADRA CSW00661, Structural Analysis for lntake Structure Trash Rake Monorail System, Rev. 1ECR 09-00685,204-45123 and 204-45611 '1O-Ton and 1S-Ton Gondensers, Rev.0EXLNOC094-PR-01, Assessment of the Oyster Creek Hardened Vent System, Rev. 0lssue Reports3305923506276250296251 38631025645374929813943876943993945376945676953452964555972528976547978288985986991 34510566231 058926105902110620051 0625001 08873511154001 12803611280421 13590011640201 16620811662201 1 668481 1 986231 1986291243623124363512765691 280593129086513178611 31 8090"1 31 8266*1 31 8288"1319776*1319787*1321771*1 321 899"1 321 996"13226731323920*1323992*13241641324254"1324795*1 31 8465*1323820.1324888*(* denotes NRC identified during this inspection)Drawinqs13432.33-EM-1, Radiation Shielding Support Reactor Cavity Drain Line Pipe Supports, Rev. 1013983-0002-E-01, Plant Process Computer System Network Block Diagram, Sh. 1 & 4, Rev. 013983-0002-E-07, Multiplexor Circuit PC6 Subnet 'B' Switch Wiring Diagram, Rev. 015595.00-EM-1, Intake Structure Modifications, Rev. 12153, Fuel Pool Cooling Filtering & Drain Piping Plans and Sections Reactor Building, Rev. 42167, HVAC Control, Mechanical Equipment & Cable Room, Rev. 5237E756, Spent Fuel Pool Cooling Flow Diagram, Rev. 533C-532-A3-1000, Pipe lntegrity Inspection Program ESW System Piping, Rev. 13D-531-22-1009, ECR 08-01040 Attachment 2, Sh. 1 , 2, & 3, Rev. 03D-532-24-001, Emergency Service Water System Pipe Restraint Modification, Rev. 03E-168-02-001 , General Arrangement lntake Structure, Rev. 104031, Intake Structure Sections and Details, Rev. 34034,Intake Structure Trash Rack and Stop Log Details, Rev. 3538361, General Erection lntake Structure, Rev. GAttachment

A-4557744, General Erection Intake Structure, Rev. DBR 2005, Reactor and Turbine Building Service Water System, Sh. 2, Rev. 105BR 2006, Turbine Building Closed Cooling Water System Flow Diagram, Sh. 5, Rev. 58BR 2010, Control and Cable Spreading Rooms HVAC Flow Diagram, Rev. 32BR 201 1, Reactor Building Ventilation, Sh. 2, Rev. 62BR 30018,4160V System One Line Diagram, Rev. 16BR 3001C, 4160V System One Line Diagram, Rev. 1BR 3005, Misc. Building 460V MCC One Line Diagram, Sh. 5, Rev. 10GE 15786350, 480V System Electrical Elementary Diagram, Sh. 41, Rev. 23GE 237E566, Reactor Protection System, Sh. 17, Rev. 3GU 3E-243-21-1000, Drywell and Torus Vacuum Relief System, Rev. 28SN 13432.19-1, Nitrogen Supply System, Sh. 1, Rev. 33U949-C-5000, Intake Structure Plan, Rev. 1Procedures205.95.0, Reactor Flood-up / Drain-down, Rev. 182400-GMM-3900.52, Inspection and Torquing of Bolted Connections, Rev. 5307, lsolation Condenser System, Rev. 116309.1, Turbine Building Closed Cooling Water System, Rev. 57310, Containment Spray System Operation, Rev. 98312.11, Nitrogen System and Containment Atmosphere Control, Rev. 40 & 41312.9, Primary Containment Control, Rev. 52 & 53344.2,lntake Trash Rake Operation, Rev. 10654.3.004, Control Room HVAC System 'A' Flow and Differential Pressure Test, Rev. 13665.5.001, Torus to Drywell Vacuum Relief Valve Leak Rate Test, Rev. 29681.4.005, Substation Tour Sheet, Rev. 17 & 18ABN-31, High Winds, Rev. 14 & 16ABN-32, Abnormal lntake Level, Rev. 18AD-AA-01, Processing of Procedures and T&RMs, Rev. 23CC-AA-1, Configuration Control, Rev. 0CC-AA-10, Configuration Control Process Description, Rev. 6CC-AA-102, Design lnput and Configuration Change lmpact Screening, Rev.22CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes, Rev.22CC-AA-107, Configuration Change Acceptance Testing Criteria, Rev. 8CC-AA-107-1001 , Post Modification Acceptance Testing, Rev. 4CC-AA-13, Margin Management, Rev. 2CC-MA-102-1001, Design Inputs and lmpact Screening - lmplementation, Rev. 9LS-AA-104, Exelon 50.59 Review Process, Rev. 6LS-AA-104-1000, Exelon 50.59 Resource Manual, Rev. 6LS-AA-120, lssue ldentification and Screening Process, Rev. 14MA-AA-743-310, Diagnostic Testing and Evaluation of Air Operated Valves, Rev. 5NRT-OC-08-0006, RTP-2000 Functional Test, Rev. 0OP-OC-108-109-1001, Severe Weather Preparation T&RM for Oyster Creek, Rev. 12Attachment

A-5Work OrdersA2149453p.2204999A226206842262069A2270528A2270529c2014505c2014848c2016918c2017315c2019099c2021995c2023483c2025008c2025252c2025388c2025389M2119029R0802188R2127181R2128890R2132325R2156772R2162526R2165640R2185877R2189853MiscellaneousAWC Flat Festoon Cable (PVC) Specifications, dated 1125112C2017315, Post-lnstallation Walkdown Checklist, dated 1 1 19110C2019099-18, SQUG Walkdown - Seismic Adequacy per ECR 09-00708, dated 11l13l}gCR HVAC System Walkdown Report, dated 8130111 & 11129111Cycle 23 Core Operating Limits Report - Oyster Creek, Rev. 5ECR OC 06-00819 Attachment F, Acceptance Test Criteria for Intake Structure Trash RakingSystem, Rev. 0First Amendment to Station Blackout Agreement between Forked River Power LLC and ExelonGeneration Company, LLC, dated 5112110GE-NE-0000-0052-5690-R0, TRACGO4 10 CFR 50.59 Evaluation Basis, Rev. 0GS 04L43B01-01E, Daqstation DX1000N General Specifications, Rev.5HVAC - Air Handling Equipment PCM Template, dated 815111f n-service Testing Bases Document (V-23-13), January 2012NEDC-33065P, Application of Stability Long Term Solution Option 2 to Oyster Creek, Rev. 0OCNGS - Relief from the Requirements of the ASME Code, Relief Request No. VR-02 for theFifth Inservice Testing Interval (TAC No. ME7618), dated 1124112OCNGS - Relief Request RP-04, Regarding SW Pump Suction Pressure Gages, and RV-51,Containment lsolation Valve Position Indication (TAC No. M84945), dated 1012102Program Health Report, NRC Generic Letter 89-13, Q4-2011Submittal of Proposed Alternative and Relief to the Requirements of 10 CFR 50.55a Concerningthe Fourth Ten-Year Interval In-service Testing Program, dated 4119102Submittal of Relief Request for the Fifth ln-service Testing lnterval (RA-11-089), dated 11117111System Health Report, Circulating Water, Q3-2011System Health Report, Control Room HVAC, Q4-2011System Health Report, Emergency Service Water, Q4-2011System Health Report, Screen Wash, Q3-2011System Health Report, Service Water, Q3-2011V-14-33 Motor Temperature Trend Data, dated 10130111 - 1122112V-14-33 Packing Leak Rate Trend Data, dated 10110111 - 1116112VM-OC-0008, Magne-Blast Circuit Breaker Vendor Manual, Rev. 14VM-OC-2888, lntake Trash Rake Installation Operation and Maintenance Manual, Rev. 1Design & Licensinq BasesNRC Regulatory Guide 1.183, Alternate Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors, July 2000OCNGS Updated Final Safety Analysis Report, Rev. 17Attachment

A-6Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 262toFacitity Operating License No. NPF-16 for the Oyster Creek Nuclear Generating Station -Application of Alternate Source Term Methodology (TAC No. MC6519), dated 4126107SDBD-OC-243, Design Basis Document for Containment System, Rev. 1TDR No. 1099, Station Blackout Evaluation Report, Rev. 4NUREG-0800 SRP-15.0.1, Radiological Consequence Analyses Using Alternate Source TermsStandard Review Plan, Rev. 0PBD-AMP-3.2.04, Oyster Creek License Renewal Project Periodic Inspection of VentilationSystems, Rev. 1TAC No. 68577, Safety Evaluation - SBO Analysis OCNGS, dated 8123191Completed Surveillance and Modification Acceptance Tests2011-002-001, ESW 2 Piping, Downstream of P-33C UT NDE Report, performed 11211112011-002-015, ESW Piping Under lntake (ES0106) UT NDE Data Report, performed 41271112011-002-016, ESW Piping Under Intake (ES0112) UT NDE Data Report, performed 41281112011-002-017, ESW Piping Under Intake (ES0212) UT NDE Data Report, performed 41291112400-SMM-3900.04 Exhibit 1, C2017315 ESW 2 Pressure Test (ASME Xl), performed 11171102400-SMM-3900.04 Exhibit 1, C2025008 ESW Tee Pressure Test (ASME Xl), performed 11241112400-SMM-3900.08 Exhibit 1, C2023483-13 General Hydrostatic Test, Initial Service Leak Test,and Pneumatic Test (ANSI 831.1), performed 11130110607.4.017, CS/ESW Pump System 2 Operability and Quarterly lnservice Test, performed 1120112636.2.012, Diesel Generator#1 Battery Service Test, performed3lTlll636.2.013, Diesel Generator #2 Battery Service Test, performed 10/9/09 & 12113110636.4.001, Diesel Generator #1 Automatic Actuation Test, performed 11124110636.4.002, Diesel Generator #2 Automatic Actuation Test, performed 11112110654.3.004, Control Room HVAC 'A' Flow and DP Test, performed 6/28/07,6123109, &6128111654.3.006, Control Room HVAC 'B' Flow and DP Test, performed 6117110654.4.003, Control Room HVAC System Operability Test, performed 1112112665.5.001, Torus to Drywell Vacuum Relief Valve Leak Rate Test, performed 1 1129110678.4.001, Primary Containment lsolation Valve Operability and lST, (V-23-13, -14, -15, -16),performed 4113111, 812111, & 10112111681.4.005, Substation Tour Sheet, performed 1125112 - 1127112C2019099-10, In-service Leak Test for 3 New Condensers Replaced by ECR 09-00708,performed 11l13l0gC2021995-07, ln-Service Leak Test for Condenser Refrigerant Pipe Assembly, Compressor,Evaporator and allAssociated Air Lines, performed 11116109ER-AA-335-015 Attachment 1, C2025008 l/'I-2, NDE Report, performed 1124111Weld Map No. 532-WM-050/0, C2017315 Pipe Weld Record Sheet, dated 11124110Weld Map No. 532-WM-060/0, C2017315 Pipe Weld Record Sheet, dated 1211110Attachment

ADAMSCFRDCDRSEDGExelonESWFRCTHVACIPIRIRCUMOVNEINRCOCNGSPARSPMTRBCCWSSCSWTBCCWTSUFSARVdcA-7LIST OF ACRONYMSAgencywide Documents Access and Management SystemCode of Federal RegulationsDirect CurrentDivision of Reactor SafetyEmergency Diesel GeneratorExelon Nuclear NortheastEmergency Service WaterForked River Combustion TurbineHeating, Ventilation and Air Conditioninglnspection Procedurelssue ReportIntelligent Remote Control UnitMotor-Operated ValveNuclear Energy InstituteNuclear Regulatory CommissionOyster Creek Nuclear Generating StationPublicly Available RecordsPost-Modification TestReactor Building Closed Cooling WaterStructure, System, and ComponentService WaterTurbine Building Closed Cooling WaterTechnical SpecificationsUpdated Final Safety Analysis ReportVolts, Direct CurrentAttachment