IR 05000352/2011007

Download: ML11353A394

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^,-a. December 19, 201.LMr. Michael J. PacilioSenior Vice President, Exelon Generation Company, LLCPresident and Chief Nuclear Officer, Exelon Nuclear4300 Winfield Rd.Warrenville, lL 60555

SUBJECT: LIMERICK GENERATING STATION - NRC EVALUATION OF CHANGES,TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONSTEAM INSPECTlON REPORT 05000352/201 1007 AND 050003531201 1007

Dear Mr. Pacilio:

On November 4, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed aninspection at the Limerick Generating Station (LGS), Units 1 and 2. The enclosed inspectionreport documents the inspection results, which were discussed on November 4,2011, withMr. Peter Gardner, Plant Manager, 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.In conducting the inspection, the team reviewed selected procedures, calculations and records,observed activities, and interviewed station personnel.This report documents one NRC-identified finding of very low safety significance (Green). Thisfinding was determined to involve a violation of NRC requirements. However, because of thevery low safety significance and because the finding was entered into your corrective actionprogram, the NRC is treating the finding as a non-cited violation (NCV), consistent withSection 2.3.2 of the NRC's Enforcement Policy. lf you contest the NCV in this report, youshould provide a response within 30 days of the date of this inspection report, with the basis foryour denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,Washington, D,C. 20555-0001, with copies to the RegionalAdministrator, Region l; theDirector, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.20555-0001; and the NRC Resident Inspector at the Limerick Generating Station. ln addition, ifyou disagree with the cross-cutting aspect of the finding in this report, you should provide aresponse within 30 days of the date of this inspection report, with the basis for yourdisagreement, to the RegionalAdministrator, Region l, and the NRC Resident Inspector at theLimerick Generating Station. In 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 theNRC'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 Nos. 50-352; 50-353License Nos. NPF-39; NPF-85

Enclosure:

I nspection Report 05000352i 201 I 007 ; 05000353/201 1 007W

Attachment:

Supplemental Informationcc w/encl: Distribution via ListServ tn 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 theNRC'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,IRAILawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetYDocket Nos. 50-352; 50-353License Nos. NPF-39: NPF-85

Enclosure:

Inspection Report 05000352/201 1007 ; 05000353/201 1 007d

Attachment:

Supplemental Informationcc w/encl: Distribution via ListServDISTRIBUTION Mencl:W. Dean, RA (RIORAMAIL Resource)D. Lew, DRA (RIORAMAIL Resource)D. Roberts, DRP (RiDRPMAIL Resource)D. Ayres, DRP (RIDRPMAlL Resource)C. Miller, DRP (RlDRSMail Resource)P. Wilson, DRS (RlDRSMail Resource)P. Krohn, DRPA. Rosebrook, DRPS. lbarrola, DRPE. Miller, DRPE. DiPaolo, DRP, SRIN. Sieller, DRP, RlN. Esch, DRP, Admin Asst.M. Franke, Rl, OEDORidsN rrPM Limerick ResourceRidsNrrDorlLpll -2 ResourceROPreports ResourceSUNSI Review Complete: (Reviewer's lnitials)DOCUMENT NAME: G:\DRS\Engineering Branch z\Mangan\LGSmodsreport2O11007.docxADAMS ML: 113534394"E=AneraOFFICE RI/DRSRI/DRPRI/DRSRI/DRSNAME KManqan/kamPKrohn/pgkCCahilUcqcLDoerflein/ltdDATE ',t211511112t6t1112t2rt112t19t11OFFICIAL U.S. NUCLEAR REGULATORY COMMISSIONREGION IDocket Nos.: 50-352,50-353License Nos.: NPF-39, NPF-85Report Nos.: 05000352/2011007 and 0500035312011007Licensee: Exelon Generation Company, LLCFacility:Limerick Generating Station, Units 1 and 2Location: Sanatoga, PA 19464Inspection Period: October 17 through November 4, 2011Inspectors: K. Mangan, Senior Reactor Inspector, Division of Reactor Safety (DRS),Team LeaderC. Williams, Reactor Inspector, DRSJ. Rady, Reactor Inspector, DRSApproved By: Lawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyEnclosure

SUMMARY OF FINDINGS

lR 0500035212011007, 0500035312011007; 1011712011-111041201 1; Limerick GeneratingStation Units 1 and 2: Evaluations of Changes, Tests, or Experiments and Permanent PlantModifications.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. One finding of very low risk significance (Green) was identified,which was considered to be a non-cited violation. The significance of most findings is indicatedby their color (Green, White, Yellow, Red) using Inspection Manual Chapter (lMC) 0609,"significance Determination Process" (SDP). Findings for which the SDP does not apply maybe Green or be assigned a severity level after NRC management review. The NRC's programfor overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006'NRC-ldentified and Self-Revealinq Findinos

Cornerstone: Mitigating Systems.

Green.

The team identified a non-cited violation of 10 CFR 50.63, "Loss of All AlternatingCurrent (AC) Power," because Exelon did not demonstrate that the alternate AC (AAC)source could provide acceptable capability to withstand a station blackout (SBO) withinthe analyzed coping timeline. Specifically, Exelon's evaluation of the Limerick GeneratingStation's excess emergency diesel generator (EDG) capacity did not analyze the effectsof the loss of an operating emergency service water (ESW) pump following a single failureon the non-blacked out unit. The loss of the ESW pump would result in loss of cooling toone of the three credited EDGs and a subsequent high temperature trip of the EDG. Theteam determined the time delay to reset this trip had not been evaluated and that Exelonhad not performed the timed test required by 10 CFR 50.63 to show that actions requiredto provide power to the blacked-out unit from the AAC could be performed within theanalysis requirements. As a result, the team concluded that Exelon did not demonstratethat the MC source would have the required availability and capability within theanalyzed timeline. Exelon entered the issue into their corrective action program forevaluation and resolution.This issue was more than minor because it is associated with the design control attributeof the Mitigating Systems cornerstone and adversely affected the cornerstone objective ofensuring the availability, reliability, and capability of systems that respond to initiatingevents to prevent undesirable consequences. The team determined the finding was ofvery low safety significance because it was a design or qualification deficiency confirmednot to result in a loss of functionality. The finding had a cross-cutting aspect in the area inthe area of Problem ldentification and Resolution, Corrective Action Program Component,because Exelon did not thoroughly evaluate problems such that resolutions addresscauses and extent of conditions and did not conduct effectiveness reviews to ensureproblems are resolved. Specifically, Exelon's recent safety evaluation did not evaluateproblems associated with a loss of an EDG due to a high temperature condition and theimpact on the SBO AAC power source availability. (lMC 0310, Aspect P.1(c)) (1R17.1b)

REPORT DETAILS

1. REACTORSAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R17 Evaluations of Chanoes. Tests. or Experiments and Permanent Plant Modifications(lP 71111.17).1 Evaluations of Chanqes. Tests. or Experiments (27 samples)a. Inspection ScopeThe team reviewed eight 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 nineteen 10 CFR 50.59 screenings for whichExelon had concluded that no safety evaluation was required. These reviews wereperformed to assess whether Exelon's threshold for performing safety evaluations wasconsistent with 10 CFR 50.59. The sample included design changes, calculations, andprocedure changes.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 screenings and applicabilitydeterminations were selected based on the safety significance, risk significance, andcomplexity of the change to the facility.In addition, the team compared Exelon's administrative procedures used to control thescreening, preparation, review, and approval of safety evaluations to the guidance inNEI 96-07 to determine whether those procedures adequately implemented therequirements of 10 CFR 50.59. The reviewed safety evaluations and screenings arelisted in the attachment.Enclosure b.2FindinqsIntroduction: The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR 50.63, "Loss of All Alternating Current Power,"because Exelon did not demonstrate that the alternate alternating current (AAC) sourceprovided the availability and capacity needed to mitigate a station blackout (SBO) withinthe analyzed one-hour coping timeline. Specifically, the team determined that Exelon'sevaluation of the availability of the non-blackout unit's excess emergency dieselgenerator (EDG) capacity did not analyze the effects of the temporary loss of one of thethree credited EDGs following an assumed single failure on the non-blacked out unit.Description: The team reviewed Exelon's 10 CFR 50.59 safety evaluation that evaluatedchanges to the emergency service water (ESW) system valve configuration. This safetyevalu-ation was performed following the issuance of unresolved item (URl) 05000352,353t2011008-01, "station Blackout Licensing Basis Assumed Alternate AC PowerSource." The URI documented the need for further evaluations to determine if the MCpower source was able to meet Limerick Generating Station's (LGS) licensing basisduring certain SBO events. The team reviewed the NRC Supplemental SafetyEvalu-ation for Station Blackout Rule (10 CFR 50.63) for Limerick Units 1 and 2, datedJune 10, 1992, which documented the NRC staffs evaluation of the LGS's loss of allalternating current power submittal. In the NRC Safety Evaluation, the staff approvedthe use of an AAC power source to supply alternating current (AC) power to the blacked-out unit. The team found that the NRC Safety Evaluation allowed the AAC source to bethe excess capacity from the non-blackout unit's EDGs. The NRC Safety Evaluationconcluded that with an assumed single failure of one of the four EDGs on the non-blackout unit, the remaining three EDGs were assumed to be available and hadsufficient capacity to shutdown both units safely.During a review of Exelon's 10 CFR 50.59 safety evaluation, the team found that,following the changes to the ESW lineup, one of the three credited EDGs would tripunder certain SBO scenarios. Specifically, in the event of a Unit 1 SBO and theassumed single failure of an EDG on Unit 2 (the non-blacked out unit), ESW cooling toone of the remaining EDGs would be lost. This would result in a subsequent high jacketwater temperature trip of the EDG. As a result, the non-blackout Unit 2 would have twooperating EDGs and require operator action to recover the third EDG in order to providethe excess capacity (three non-blackout unit EDGs) assumed in the NRC SafetyEvaluation. The team noted that Exelon had determined that power to the affected ESWpump would be restored by providing power to the SBO unit's 4kV bus from a non-blackout unit 4kV bus via a safety bus cross-connection in accordance with the SBOemergency operation procedures. Exelon concluded that once power was reslored toan ESW pump, ESW flow would be restored allowing for recovery of the third EDG,therefore, no change to the license was required.The team identified that Exelon's evaluation of the loss of the third EDG due to a highjacket water temperature trip did not consider the time required for the temperature tripio reset. The team found that Exelon had assumed that when ESW was restored to theEDG the temperature switch would quickly reset. However, because the jacket waterpump would not be operating the team questioned how long it would take to cool theEnclosure 3jacket water system in order to reset the trip and allow the EDG to be started. Inresponse to the teams questions, Exelon performed thermalcalculations to determinethe time required to cool down sufficiently; however, because there was a large variancein the time based on calculation assumptions the team concluded that the calculationsdid not demonstrate that the third EDG could be restored within the analyzed one-hourcoping timeline. Additionally, the team could not determine the actual time available toallow for recovery of the EDG because Exelon did not have records of a demonstrationthat recorded the time required to power the blacked out unit from the MC source.Therefore, the team concluded the Limerick Generating Station's MC power was not inconformance with the analysis and did not meet SBO requirements. The team notedthat although LGS was not in conformance with the analysis submitted to the NRC todemonstrate compliance with the SBO Rule, LGS did have procedures in place andadditional equipment capacity (EDG and DC battery) not credited in the analysis thatwould allow the unit to cope with a station blackout until the third EDG could berestarted. Exelon entered these issues into their corrective action program forevaluation and resolution under CR 01288965.Analysis: The team determined that the failure to verify the AAC source would beavailable within the analyzed timeframe during an SBO event was a performancedeficiency. Specifically, Exelon's 10 CFR 50.59 safety evaluation did not include acomplete evaluation of the affects of a high temperature trip on a non-blackout EDG withrespect to the non-blackout unit's ability to provide the AAC source within the analysistimeline assumed in the NRC Safety Evaluation. The team concluded that thisperformance deficiency was reasonably within Exelon's ability to foresee and prevent.This issue was more than minor because it was similar to NRC Inspection ManualChapter (lMC) 0612, Appendix E, "Examples of Minor lssues," Example 3.j, in that as aresult of this deficiency; the team had a reasonable doubt of operability with respect tothe MC power source capacity to recover from an SBO. ln addition, the finding wasassociated with the design control attribute of the Mitigating Systems cornerstone andadversely affected the cornerstone objective of ensuring the availability, capability, andreliability of systems that respond to initiating events to prevent undesirableconsequences.The team performed a Phase 1 SDP screening, in accordance with NRC IMC 0609,Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," anddetermined the finding was of very low safety significance (Green) because it was adesign or qualification deficiency confirmed not to result in a loss of functionality of theequipment. Specifically, a single failure of an EDG on the non-blacked out unit did notneed to be assumed per the SDP. The team identified a cross-cutting aspect associatedwith the finding in the area of Problem ldentification and Resolution, Corrective ActionProgram Component, because Exelon did not thoroughly evaluate problems such thatresolutions address causes and extent-of-conditions and did not conduct effectivenessreviews to ensure problems are resolved. Specifically, Exelon's recent 10 CFR 50.59safety evaluation did not evaluate problems associated with a loss of an EDG due to ahigh temperature condition and the impact on the SBO AAC availability. (lMC 0310,Aspect P.1(c))Enclosure

.2.2.1 a.4Enforcement. 10 CFR 50.63, "Loss of AllAlternating Current Power," requires that aplant be able to withstand for a specified duration and recover from an SBO. An AACpower source constitutes the acceptable capability to withstand an SBO provided ananalysis is performed which demonstrates that the plant has this capability from onset ofthe SBO until the MC source and required shutdown equipment are started and linedup to operate. In addition, the time required for startup and alignment of the AAC powersource and this equipment shall be demonstrated by test. Contrary to the above, afterchanges were made to the ESW lineup in October 12,2001, Exelon did not determinewhether the non-blackout unit's EDGS were capable of providing the necessary excesscapacity within the analyzed one hour coping timeframe. Because this finding was ofvery low safety significance and was entered into the corrective action program(lR 01288965), this violation was treated as a non-cited violation, consistent withSection 2.3.2 of the NRC Enforcement Policy. (NCV 05000352/2011007-01, Failure toVerify Alternate AG Source Capability to Recover from Station Blackout)Permanent Plant Modifications (10 samples)Use of Ultralow Sulfur Diesel Fuelfor the Emerqencv Diesel GeneratorsInspection ScopeThe team reviewed a modification (07-00049) that evaluated the acceptability oftransitioning from 5500 (500 ppm sulfur) low sulfur dieselfuel oil to 315 (15 ppm sulfur)ultra-low sulfur diesel (ULSD) fuel oilfor use in the EDGs. The transition was made tomeet Environmental Protection Agency rules and standards. Exelon performed themodification in order to evaluate the effect ULSD fuel oil would have on the performancecapability of the EDGs, and to verify that the design and licensing bases for LGS werenot impacted by the use of the ULSD fuel oil. Additionally, the evaluation determined theactions required by the site to support the fuel change.The team reviewed Exelon's evaluations for use of the ULSD fuel oil, as well industryoperating experience, to determine if compatibility issues with ULSD fuel oil wereappropriately addressed. The team reviewed the revised diesel storage and fuel oilconsumption calculations, and discussed the calculations with the responsible designengineers to determine if the calculation assumptions were appropriate and the requiredvolume of ULSD fuel oil was in accordance with the licensing requirements of the plant.The team also reviewed fuel oil procurement and sample procedures, and receiptrecords to determine if Exelon was appropriately monitoring ULSD fueloil parameters.Finally, the team reviewed condition reports (CRs) and EDG testing records to verify thatEDG performance was not impacted by the fuel oil change. The 10 CFR 50.59screening determination associated with this modification was also reviewed asdescribed in section 1R17.1 of this report. Documents reviewed are listed in theattachment.FindinasNo findings were identified.b.Enclosure

5.2,2 lr&rdification tglLhe U1 Hioh Pressure Coolant Iniection Booster Pump 9ouplincla. Inspection ScopeThe team reviewed a modification (10-00126) to the high pressure coolant injection(HPCI) booster pump coupling. The coupling connects the rotating assemblies of theHPCI booster pump and the HPCI main pump. Exelon performed this modification toallow for improved and easier disassembly of the coupling for maintenance activities.The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the HPCI pump had not been degraded by the modification.The team interviewed Exelon's engineering staff and reviewed the vendor technicalevaluation to determine if the coupling modification had impacted the pump or couplingperformance. The associated work order instructions and documentation were reviewedto verify that maintenance personnel implemented the modification as designed. Theteam also walked down the HPCI booster pump and HPCI booster pump coupling todetermine if the maintenance activities were performed in accordance with themodification procedures. Finally, the team reviewed surveillance test results todetermine if the HPCI pump performance had been adversely impacted. The10 CFR 50.59 screening determination associated with this modification was alsoreviewed as described in section 1R17,1 of this report. Documents reviewed are listedin the attachment.b. FindinqsNo findings were identified..2.3 Replacement of the 2B-E205 Residu4l-.j-leat Removal Heat Exchanoera. Inspection ScopeThe team reviewed a modification (09-00333)that replaced the 2B-E205 residual heatremoval (RHR) heat exchanger. The RHR heat exchanger removes heat from thereactor after plant shutdown, and removes heat from the primary containment duringcertain design basis accidents. Exelon implemented this modification to replace theexisting RHR heat exchanger that was approaching design limits. The new heatexchanger was selected by Exelon because it was dimensionally similar to the existingheat exihanger, and was built with alloy steel tubes which had improved corrosionresistance and performance for the component'The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the RHR heat exchanger and associated system had not beendegraded by the material change to the heat exchanger tubes or the heat exchangerinstallation. The team interviewed design engineers and reviewed vendor data,calculations, and evaluations to determine if the capacity of the new heat exchanger metthe design and licensing requirements. Additionally, the team reviewed post-modification testing (PMT) results, and associated maintenance work orders to verifythat the heat exchanger replacement modification was appropriately implemented.Enclosure 6Finally, the team walked down the heat exchanger with the system engineer to verify themaintenance activities were performed as described in the modification package. The10 CFR 50.59 screening determination associated with this modification was alsoreviewed as described in section 1R17.1 of this report. Documents reviewed are listedin the attachment.b. FindinosNo findings were identified..2.4 Modificatiofr of Residual Heat Removal Service Water'B' Return Loop Pipinoa. Inspection ScopeThe team reviewed a modification (09-00134) that installed a drain valve assembly in theRHR service water (RHRSW) return loop piping. The RHRSW return loop piping returnsRHRSW to the spray pond. Exelon performed the installation of the drain valveassembly to repair a PiPe flaw.The team reviewed the modification to determine if the design basis, licensing basis, orperformance capability of the return line had been degraded by the modification' Theteam interviewed design and non-destructive testing engineers, and reviewedevaluations, non-destructive testing results, PMT results, and associated maintenancework orders. This review was performed to verify the flaw was repaired by theinstallation of the assembly, the repair met the requirements of the American Society ofMechanical Engineers (ASME) Code, and that the drain valve modification wasappropriately implemented. The team also verified that the drain valve assemblyspecifications, associated procedures, and drawings had been updated. Finally, theteam walked down the drain valve with the system engineer to verify the maintenanceactivities were performed in accordance with the work order. The 10 CFR 50'59screening determination associated with this modification was also reviewed asdescribed in section 1R17.1 of this report. Documents reviewed are listed in theattachment.b. FindinqsNo findings were identified..2.5 Unit 2 Uleasurement Uncertaintv Recapture Power Uprate Leadinq Edqe Flow Metera. Inspection ScopeThe team reviewed a modification (09-00097) that installed the Leading Edge FlowMeter (LEFM) CheckPlus System in Unit 2's three main feedwater piping return headers.The modification was performed to reduce the two percent uncertainty margin asoriginally required by 10 CFR Part 50, Appendix K. Feedwater flow signals frominsialled flow venturis had been used for determining core thermal power. The LEFMmodification was performed to provide feedwater mass flow signals as the primary input b.a..2.67to determine core thermal power. The modification included installation of a meteringspool piece that consisted of 16 ultrasonic transducers, a common pressure tap for twonew pressure transmitters, and a thermowell for the dual element resistancetemperature detector.The team reviewed the modification to determine if the design basis, licensing basis, andperformance capability of the feedwater flow measurement system had been degradedby the modification. The team reviewed calculations and technical evaluations, andinterviewed system and design engineers to assess whether the modification wasconsistent with design assumptions, Replacement components and materials werereviewed to ensure that the modification conformed to the design specifications for thefeedwater system. The team also reviewed design assumptions and supportinguncertainty calculations to evaluate whether they were technically appropriate andconsistent with the UFSAR, and to ensure design limits were not exceeded. The teamreviewed the post-modification testing and vendor commissioning documents to verifyproper operation of the system. Finally, the team reviewed CRs associated with thesystem installation to verify that deficiencies were appropriately identified and corrected.The 10 CFR 50.59 screening determination associated with this modification was alsoreviewed as described in section 1 R17.1 of this report. Documents reviewed are listedin the attachment.FindinosNo findings were identified.Technical Specification 3.8.1 lntent Chanoed Without Prior NRC Approvallnspection ScopeThe team reviewed a modification (09-00284) that returned wording in the TS Bases314.8.1 document to the wording used prior to implementation of modification 99-00682.ln 1999, Exelon implemented 99-00682 which changed the TS Bases 314.8.1wording tostate that only three out of four 4 kV emergency buses were required to be electricallyconnected to offsite power to maintain the operability of the offsite power sources. In aprevious inspection report, NRC inspectors determined that this change to the TS Baseschanged the intent of the associated TS 3/4.8.1 and issued a Severity Level lV non-citedviolation of 10 CFR 50.59, "Changes, Tests, and Experiments," (NCV 05000352,353/2009002-02) for failing to obtain a TS license amendment prior to changing thewording. Modification 09-00284 changed the TS Bases wording to require allfouremergency buses be connected to offsite power in order to consider offsite power to beoperable.The team reviewed the modification to verify that the approved design and licensingbases had not been changed by the modification. The team noted the modification wasonly a change to the TS Bases document and did not require any plant equipmentchanges. The team reviewed the design and licensing bases assumptions to evaluatewhether the modification was appropriate and consistent with the UFSAR. Also, theteam reviewed CRs associated with the original modification and associated violation toEnclosure

.2.7 Idetermine if the deficiencies identified were appropriately corrected. The 10 CFR 50.59screening determination associated with this modification was also reviewed asdescribed in section 1R17.1 of this report. Documents reviewed are listed in theattachment.FindinqsNo findings were identified.Multiple Spurious Operation: Generate-2R11 ECR for Mods to Core Sprav and ResidualHeat Removal Check ValvesInspection ScopeThe team reviewed a modification (10-00347) that changed the test circuit wiring for thecore spray (CS) and RHR testable check valves in order to prevent the valves fromspuriously opening, due to a hot short, during a postulated fire scenario. Themodification was performed because a hot short could bypass the testable check valvepushbutton switches in the main control room and cause the testable check valves tospuriously open. The modification added wiring to the test circuit of each testable checkvalve in order to cause a short to ground in the event that a postulated hot shortoccurred during a fire.The team reviewed the design basis, licensing basis, and performance capability of theCS and RHR testable check valves. The team evaluated the modification to ensure itwas consistent with requirements in the design and licensing bases, and that thecomponents had not been degraded. The team reviewed technical evaluations todetermine whether the modification was consistent with design assumptions for valveoperation. Electrical elementary wiring diagrams were reviewed to verify that thetestable check valves were not adversely affected by the modification, and replacementmaterials were reviewed to ensure that they conformed to the system designspecifications. The team also verified selected drawings and procedures were properlyupdated for the new equipment configuration. Additionally, the team reviewed the post-modification testing performed to verify proper operation of the CS and RHR testablecheck valves to determine if the results were satisfactory. Finally, the team conductedinterviews with engineering staff to determine if the testable check valves functioned inaccordance with the design assumptions, and if the modification corrected the previouslyidentified problem. The 10 CFR 50.59 screening determination associated with thismodification was also reviewed as described in section 1 R17.1 of this report.Documents reviewed are listed in the attachment.FindinosNo findings were identified,b.Enclosure

a..2.89Incorporated Shroud Evaluation into Desiqn BasisInspection ScopeThe team reviewed a modification (09-00035) which was an engineering evaluationperformed to re-evaluate the structural integrity of the Unit 2 core shroud. Specifically,the modification evaluated the welds on the core shroud to determine an acceptable timeinterval for in-service inspections of the welds that connected the sections of the shroud.The evaluation determined how many cycles of operation could occur before re-inspection of the core shroud welds would be required to validate the assumptions in themethodology used to determine the structural integrity of the weld. To perform thisevaluation Exelon utilized the RAMA (Radiation Analysis Modeling Application) Codemethodology.The team reviewed the modification to determine if the design and licensing basesrequirements for the Unit 2 core shroud welds were met. The team assessed if themethodology was in accordance with the guidance of Regulatory Guide 1

.190 andevaluated the basis for the inputs into the code. The team also determined if Exelonsatisfactorily evaluated the results of the evaluation in order to determine the appropriateshroud weld inspection interval. Additionally, the 10 CFR 50.59 screening determinationassociated with this modification was reviewed as described in section 1R17.1 of thisreport. The documents reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.9 Total Inteqrated Dose (TlD) Evaluation for Drvwell Coatinos (paint)a. Inspection ScopeThe team reviewed a modification (11-00122) which revised calculation LM-0675 - TIDEvaluation for Drywell Coatings. The calculation determined the total dose to qualifiedcoatings inside the drywell. For the new calculation, Exelon changed the evaluationmethodology from an infinite cloud evaluation to the semi-infinite cloud model because itwas determined that the infinite cloud model overestimated the total dose to thecoatings. Additionally, the revision to the calculation was based on a 60 year expecteddose to the coatings.The team reviewed the modification to determine if the design and licensing bases forthe evaluation of the drywell coating systems remained valid. The team reviewed thecalculation to verify the assumptions used were valid and the coatings had beenqualified to receive the doses determined by the calculation without failing. Finally, theteam determined if the new methodology was an acceptable methodology fordetermining coating dose and had been reviewed by the NRC. Additionally, the10 CFR 50.59 screening determination associated with this modification was reviewedas described in section 1R17.1 of this report. The documents reviewed are listed in theattachment.Enclosure

I10b. FindinosNo findings were identified..2.10 Setpoint Chanoe for Temperature Indicatinq Switch (Tl5)-025-101/201a. Inspection ScopeThe team reviewed modification (09-00551) which evaluated the permanent change tothe TIS-025-1011201. The TIS actuates based on the delta{emperature (delta-T) HPCIroom trip set-point. The delta-T trip provides a signal to isolate the HPCI steam piping inthe event of a design basis steam leak in the room. The set-point change was madebecause the previous setpoint was determined to be non-conservative. Duringrevisions of various calculations, Exelon determined that room temperature following asteam line break would not exceed the previous trip set-point.The team reviewed the modification to verify that the design and licensing bases of theisolation system had not been degraded by the set-point modification. The teamdetermined if Exelon had evaluated the impact of the delta-T trip set-point andappropriately calculated the new set-points. The team also verified the calibrationprocedures were updated for the revised set-points. Finally, the team reviewed thetechnical specifications to verify that limits in the TS had been appropriately revised andthat no TS violations had occurred. Additionally, 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.4.

4OA2 ldentification and Resolution of Problems (lP 71152)a. lnspection ScopeThe team reviewed a sample of CRs 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 or completed corrective actions were appropriate. ln addition, the teamreviewed CRs written on issues identified during the inspection to verify adequateproblem identification and incorporation of the problem into the corrective action system.The CRs reviewed are listed in the attachment.b. FindinqsNo findings were identified.Enclosure

114OA4 Othera.Unresolved ltem 05000352.353i2011008-01 - Station Blackout Licensino BasisAssumed Alternde AC Power Source (Closed)The team reviewed URI 05000352,35312011008-01, "station Blackout Licensing BasisAssumed Alternate AC Power Source." The URI was opened to evaluate if the changesthat Exelon performed on the ESW system lineup impacted the SBO licensing basis.Specifically, NRC inspectors determined that following a worst case single failure on thenon-blacked out unit (including the single failure of the EDG assumed in the NRC SafetyEvaluation on SBO), the third EDG credited in the SBO analysis would trip on hightemperature and questioned whether this would be considered a malfunction of the EDGand, therefore, the EDG could not be credited under the current licensing basis. Exelonacknowledged that the EDG may trip on high temperature but believed that the EDGcould be retovered and, therefore, be credited as one of the three EDGs required by thelicensing basis for SBO.The team reviewed the NRC Safety Evaluation for Limerick Generating Station, datedJune 10, 1992, to determine the EDG configurations required to mitigate an SBO. Theteam determined that the NRC's approval of Exelon's strategy to meet the SBO rule wasbased on excess capacity from the non-blacked out unit's EDGs. During a single unitSBO the non-blacked oui unit was assumed to have three of their four EDGs available'This scenario was based on a common cause failure of all EDGs on the blacked out unitand an assumed single active failure of one EDG on the non-blacked out unit. The teamfound that the non-blacked out unit required the capacity of more than one but less thantwo EDGs to achieve safe shutdown. The analysis credited the excess capacity of thethree remaining EDGs to be available to safely shutdown the unit affected by the stationblackout, during the four hour coping period of the SBO. Additionally, the teamdetermined thai the NRC Safety Evaluation stated that the excess capacity would beavailable within one hour of the start of the SBO.The team reviewed Exelon's SBO procedures, electrical configurations, and emergencyservice water alignments to determine if the systems were able to be cross{ied, if plantprocedures correctly directed operators to complete the alignment, and if the excesscapacity would be available within one hour. Specifically, the team reviewed the eventscenarib where the third EDG tripped due to the worst case single failure and operatoraction was required to recover the EDG. The team concluded that if the excess capacitywas able to be placed on the blacked out unit's vital buses within one hour, the LGSlicensing basis was met. This unresolved item is closed.FindinqsOne finding was identified. See Section 1R17.1.b' for details'b.Enclosure 124OAO Meetinqs. includinq ExitThe team presented the inspection results to Mr. Peter Gardner, Plant Manager, andother members of Exelon's staff at an exit meeting on November 4,2011. The teamreturned the proprietary information reviewed during the inspection and verified that thisreport does not contain proprietary information.Enclosure Licensee PersonnelP. GardnerD. DoranW. LewisR. GeorgeR. HardingA. LambertJ, MitturaM. GiftK. CollierN. RoyR. SchwabE. HostermanL. HemlerJ. BergA-1ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Plant ManagerDirector of EngineeringSenior Manager Engineering DesignManager Electrical DesignRegulatory AssuranceDesign EngineerDesign EngineerDesign EngineerDesign EngineerDesign EngineerDesign EngineerDesign EngineerSystem EngineerSystem Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSEDOpened and

Closed

050003s2/2011007-01Closed05000352,35312011008-01 URINCV Failure to Evaluate Station Blackout Timeline forEDG Availability (section 1 R1 7.1 b)Station Blackout Licensing Basis AssumedAlternate AC Power Source (section 4OA5)

LIST OF DOCUMENTS REVIEWED

10

CFR 50.59 EvaluationsLG2009E001, 'C' SLCS Pump Control Switch Modification to Inhibit Automatic Pump Start onUnits 1 and 2, Rev. 0LG2009E002, Provide an Alternate Means of Monitoring Reactor Well Seals Leak, Rev. 1LG2010E002, Suppression Pool Gross Input Leak Rate Determination, Unit 1 Rev. 15 and Unit2Rev. 18LG2010E003, Modify Select MOV Circuits to Prevent Spurious Operations during Postulated HotShort Fire Scenarios, Rev. 0LG2011E001, ECR

LG 10-00103 use of TMCGO4P Version 4.2.60.3, Rev' 01G2011E002, Modify Select MOV Circuits to Prevent Spurious Operations during Postulated HotShort Fire Scenarios, Rev. 0Attachment

A-2LG201 1E003, Reactor Recirculation M/G Set Replacement with Adjustable Speed Drive (ASD)Units, Rev.01G201 1 E004, Assessment of the Affect of

ECR 01-00816 on Station Black Out Coping, Rev. 010

CFR 50.59 Screened-out EvaluationsLG2009S004, Incorporating Shroud Evaluation into Design Basis, Rev. 0LG2009S010, UFSAR Chapter I Changes - Spare Safeguard Transformer, Rev. 01G2009S025, Revised Calcs M-81-1 0, M'81-27 and M-81-28, Rev' 0LG2009S026, UFSAR Section 8.1 Revision (Offsite Sources), Rev' 0LG2009S032, Tech Spec 3.8.1 Intent Changed without Prior NRC Approval, Rev. 0LG2009S036, ESW Loop'A' Flow Balance, Rev. 01G2009S054, Permanent Setpoint Change for

TIS-025-10112018&D, Rev. 1LG2010S016, New Allowable Total Connection Resistances for Station Batteries, Revs. 36and 38LG2010S066, Leading Edge Flow Meter CheckPlus Installation, Rev. 11G20105073, Installation of Support Equipment for CRE Connections in Unit 2 AC/BD RHRRooms, Rev.0LG2011S001, Fluence Calculation Incorporation, Rev' 0LG201 15022, MSOPS: Generate ECR for 1R14 DC Bucket Mods, Rev. 0LG20'11S028, LGS RHR and Core Spray Loop'A' Unit 1 Testable Check Valve/Bypass ValveCircuit Modification, Rev. 0LG201 1S035, Prepare ECR for Revision of TID Calc for Drywell Coatings, Rev. 0Modification PackaqesOtOOglO, Operability Determination and NCR for ESW and Emergency D/G's, Rev. 007-00049, Use of Ultra-Low Sulfur Diesel Fuel, Rev' 2*09-00035, Incorporated Shroud Evaluation into Design Basis, Rev' 009-00097, Unit 2 Measurement Uncertainty Recapture (MUR) Power Uprate Leading Edge FlowMeter (LEFM), Rev. 109-00134,

HBC-507-01 SWz Piping Modification, Rev.2*09-00284, Tech Spec 3.8.1.1 Intent Changed Without Prior NRC Approval, Rev. 0*09-00333, Replacement of 28 RHR Heat Exchanger, Rev. 8*09-00485, Compartment Temperature Transients for Steam and Water Leaks, Rev. 009-00551, Permanent Setpoint change for

TIS-025-1011201 B&D, Rev. 210-00126, HPCI Booster Pump Coupling Modification, Rev. 0*10-00347, Multiple Spurious Operations: Generate 2R11 ECR for Mods to Core Spray andResidual Heat Removal Check Valves, Rev' 233311-00122,T\D Evaluation for Drywell Coatings, Rev. 0(* designates a Modification and 10

CFR 50.59 screen-out evaluation sample)Calculations. Analvsis. and Evaluations0000-0125-5142, HPCI Speed Increase Evaluation, Rev'

0364586, Ultra-Low Sulfur Diesel Fuel Evaluation, dated 21161076380E.07, Diesel Generator Loading (Steady State), Rev' 12ER-LG-331, Augmented Inspection Program - Aug 20 Core Shroud Inspection, Rev. 1HBC-507-H002, Temporary Brace at Pipe Support, Rev. 0LE-0052, Class 1E Battery Load Duty Cycle Determination, Rev. 12LE-0111, Xformer Inrush and Motor Starting Current Transients during EDG Cross-Tie, Rev. 0Attachment

A-3LE-0114, Reactor Core Thermal Power Uncertainty Calculation Unit 2, Rev. 1LEAE-MUR-0003, Bounding Uncertainty Analysis for Thermal Power Determination at LimerickUnit 2 using LEFM CheckPlus System, Rev. 0LG-MISC-02, PRA Sensitivity Study for the Potential lmpacts of Increasing the Suppression PoolCooling Run Time, Rev. 0LG-PRA-010, LGS PRA Data Notebook Volume 1, Rev. 1LM-0007, Diesel Generator Fuel Oil Consumption, Rev. 4LM-0052, Differential Pressure Calculations for MOVs in the HPCI System, Rev. 7LM-014, Determine Sizing and Configuration of LGS Unit 1 RHR Test Return Line, Rev. 1LM-0663, Diesel Generator Day Tank Minimum Level, Rev. 2LM-0675, TID Evaluation for Drywell Coatings, Rev. 0LM-280, Radiation Through Bioshield Walland Streaming Through Bioshield Penetration,dated 312193M-11-32, Heat Exchangers Input Data for Computer Performance Program, Rev. 5M-55-03, HPCI Steam Supply Pressure Drop, Rev. 6M-55-20, HPCI Pump Discharge Maximum Pressure, Rev. 5M-81-10, Spray Pond Pump Facility Ventilation Requirements, Rev' 4M-81-27, Spray Pond Pump Station - Minimum Temperature in the Small Room, Rev. 3M-81-28, Spray Pond Pump Structure Temperature-Time Curve After a LOCA/LOOP, Rev. 2NED

C-32847P, ARTS Flow-Dependent Limits with Turbine Bypass Valve Out of Service forPeach Bottom Atomic Power Station and Limerick Generating Station, dated 6/98Condition Reports00534749006562690067383200691 5750072347200885528009052200104757601 1 388610128242501285226"01285263*01 286023.01286047.01 288965.(* denotes NRC identified during this inspection)Drawinqs8031-M-1 1, Sht. 1, Emergency Service Water, Rev. 708031-M-12, Shts. 1-2, Residual Heat Removal Service Water, Revs' 70 andT8031-M-51 , Shts. 1-8, Residual Heat Removal, Revs. 65, 66, 67, 66, 30, 23, 21, and 258031-M-53, Sht. 3, P&lD Fuel Pool Cooling and Cleanup, Rev' 168031-M-56, HPCI Pumpffurbine Unit 1, Rev.40CA34471, Forged Steel Maximum Bore Hub Puller Holes, Rev. 2M-1-E1

1-1040-E-032, Sh. 1 , Elementary Diagram Residual Heat Removal System, Rev. 13M-1-E11-1040-E-035, Sh. 1, Elementary Diagram Residual Heat Removal System, Rev. IProceduresA-C-134, Control of Hazards Barriers, Rev. 4ARC-BOP-20C222, D3 Reactor Well Seal, Rev. 0ARC-MCR-107, A-3 Alarm Response Card, Rev. 1ARC-MCR-21?,15 Fuel Pool Storage Hi/Lo Level, Rev. 1E-1, Loss of All AC Power (Station Blackout), Rev. 40E-10/20, Loss of Offsite Power, Rev. 44ER-AA-340, Generic Letter 89-13 Program lmplementing Procedure, Rev' 6Attachment

A-4lC-1 1.00388, Calibration of HPCI Turbine Governor Control System for the Limerick GeneratingStation, Rev.8LS-M-104, Exelon 50.59 Review Process, Rev. 6M-093-004, 480 VAC MCC Breaker Assembly and Cubicle Terminal Maintenance, Rev. 10OS12.1 .A, Alignment for Normal Operation of the Residual Heat Removal Service Water Systemfor Loop'B', Rev. 20P-305, Welding and Non-Destructive Testing Requirements for Field Erected Piping, Rev. 27PES-P-006, Diesel Fuel Oil, Rev. 8RT-2-012-391-2, 2B-E-205 RHR Heat Exchanger Transfer Test, Rev' 6RT-6-041-490-1, Suppression Pool Gross lnput Leak Rate Determination, Rev. 16RT-6-041-490-2, Suppression Pool Gross lnput Leak Rate Determination, Rev. 19551.8.A, Suppression Pool Cooling Operation and Level Control, Rev' 42S53.0.A, Normal Makeup/Response to Low Level in Fuel Storage Pool or Reactor Well, Rev. 22S92.3.N, Receiving Diesel Fuel Oil Delivery, Rev. 36ST-5-020-810-0, Diesel Generator Fuel Oil Receipt Analysis, Rev. 28ST-5-020-81 1-1, Diesel Generator Fuel Oil Post Receipt Analysis, Rev' 143T-6-055-230-112, HPCI Pump, Valve, and Flow Test, Rev. 76Work Ordersc0231554c0234023c0235918c0235919c0235924R1010379R1023526R1030435R1069699R1075525R1101708R1108012R1141314MiscellaneousnHtSt UtO1.Z, Protective Coatings (Paints) for Light Water Nuclear Reactor ContainmentFacilities, dated 1972ASTM D4082-10, Standard Test Method for Effects of Gamma Radiation on Coatings for Use inNuclear Power PlantsBWRVIP-114-A, BWR Vessel and Internals Project - RAMA Fluence Methodology TheoryManual, dated 6/09GE-NE-0000-0052-5690, TRACG04 DIVOM 10

CFR 50.59 Evaluation Basis, Rev. 0GE-NE-0000-0115-7421, TRACGO4P (Version 4.2.60.3) DIVOM 10

CFR 50.59 Evaluation Basis,Rev.0GNF-S-0000-0109-4007, TRACGO4P Error Correction 10

CFR 50.59 Evaluation Basis, Rev. 1NEI 96-07, Nuclear Energy Institute Guidelines for 10

CFR 50.59 lmplementation, Rev. 1NRC Generic Letter 1998-13, Service Water System Problems Affecting Safety-RelatedEquipment, dated 7 118189NRC Information Notice 1987-10, Potential for Water Hammer during Restart of Residual HeatRemoval Pumps, dated 2111187NRC lnformation Notice2AQ6-22, New Ultra-Low-Sulfur Diesel Fuel Oilcould Adversely lmpactDiesel Engine Performance, dated 10112106NRC lnformation Notice 2010-17, Common Cause Failure of BWR Recirculation Pumps withVariable Speed Drives, dated 9110110NUMARC 87-00, Guidelines and Technical Bases for NUMARC lnitiatives Addressing StationBlackout at Light Water Reactors, Rev. 1NUREG-0588, lnterim Staff Position on Environmental Qualification of Safety-Related ElectricalEquipment, Rev. 1Attachment

A-5TP24, Caldon Ultrasonics Verification and Validation Data Package Documents Vol. ll, Rev. 21TSH-GA-1 1 0A-1, lnstrument Calibration SheetSg rvei I la nce jr nd-Mod ification Acceptance Tests0000{129*688-fti, Summary of GEH Transient Anticipated Operational Occurrences (AOO)with Respect to ASD Modification in LGS Units 1 and 2, Rev. 1A5E02029143A, High Availability VFD Drive: Failure Modes Effects Analysis and ProbabilisticRisk Assessment, Rev.

AEER-790, An Evaluation of the lmpact of 55 Tube Permutit Flow Conditions on the Meter Factor ofan LEFM CheckPlus System, Rev. 1ER-797, Meter Factor Calculation and Accuracy Assessment for Limerick Unit 2, Rev. 0FCDP-197, LEFM CheckPlus 2000FC Flow Measurement System Field Commissioning DataPackage, Rev. 0MAT 09-00097-1, Unit 2 LEFM Modification Acceptance Test, Rev. 0ST-4-015-490-2, Reactor Well Seals Leak Test, performed 3122109ST-5-020,810-0, Diesel Generator Fuel Oil Receipt Analysis, performed

819111 and 8119111ST-S-020-811-1, Diesel Generator Fuel Oil Post Receipt Analysis, performed 10/6i10 and81111115T-6-051-232-2,'B'RHR Pump, Valve, and Flow Test, performed 41191115T-6-092-111-1, Diesel Generator 24-Hour Endurance Test, performed 9129110Desion & Licensino BasesLetter from Philadeipnii Etectric Company to NRC, Limerick Generating Station, Units 1 and 210

CFR 50.63, "Loss of All Alternating Current Power" Supplemental Information,dated 4lgl90Letter from Philadelphia Electric Company to NRC, Limerick Generating Station, Units 1 and 210

CFR 50.63, "Loss of All Alternating Current Power" Response to NRC SafetyEvaluation, dated 91 4191Letter from USNRC to EPRI, US Nuclear Regulatory Commission Approval Letter forBWRVIP-1 17-A, '3f,MA Fluence Methodology for Plant Application - Susquehanna Unit2 Surveillance Capsule Fluence Evaluation for Cycles 1-5", dated 4118111Limerick Generating Station Units 1 and 2 - Shutdown of the Non Blacked-out Unit with 2 Dieselsin the First Hour Following an SBO, dated 1012112011Limerick Generating Station Updated Final Safety Accident Report, Rev. 15L-S-07, Diesel Generator and Auxiliary Systems DBD, Rev. 12Safety Evaluation by The Office of Nuclear Reactor Regulation - Station Blackout SafetyEvaluation Philadelphia Electric Company, Limerick Generating Station Units 1 and 2,dated 6110192SAIC-91/6651, Technical Evaluation Report Limerick Generating Station, Units 1 and 2 StationBlackout Evaluation, dated

318191

A-6

LIST OF ACRONYMS

ADAMS Agencywide Documents Access and Management SystemMC Alternate Alternating CurrentAC Alternating CurrentASD Adjustable Speed DriveASME American Society of Mechanical EngineersCFR Code of Federal RegulationsCR Condition ReportsCS Core SprayDRS Division of Reactor SafetyEDG Emergency Diesel GeneratorESW Emergency Service WaterHPCI High Pressure Coolant InjectionIMC Inspection Manual ChapterLEFM Leading Edge Flow MeterLGS Limerick Generating StationNCV Non-Cited ViolationNEI Nuclear Energy InstituteNRC Nuclear Regulatory CommissionPARS Publicly Available RecordsPMT Post Modification Testppm Parts Per MillionRHR Residual Heat RemovalRHRSW Residual Heat RemovalService WaterSBO Station BlackoutSDP Significance Determination ProcessTS Technical SpecificationsUFSAR Updated Final Safety Analysis ReportULSD Ultra Low Sulfur DieselURI Un-resolved ltemAttachment