IR 05000333/2012007

IR 05000333-12-007; 4/16/2012 - 5/3/2012; James A. FitzPatrick Nuclear Power Plant (Fitzpatrick); Engineering Specialist Plant Modifications Inspection

ML12166A406
Person / Time
Site:	FitzPatrick
Issue date:	06/14/2012
From:	Doerflein L T Engineering Region 1 Branch 2
To:	Michael Colomb Entergy Nuclear Northeast
References
IR-12-007
Download: ML12166A406 (24)

Text

UNITED STATESNUCLEAR REGULATORY COMMISSIONREGION I21OO RENAISSANCE BOULEVARD. SUITE 1OOKING OF PRUSSIA, PENNSYLVANIA 1940S2713June 14, 20L2Mr. Michael ColombSite Vice PresidentEntergy Nuclear NortheastJames A. FitzPatrick Nuclear Power PlantP. O. Box 110Lycoming, NY 13093

SUBJECT: JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC EVALUATION OFCHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANTMODIFICATIONS TEAM INSPECTION REPORT O5OOO333/2012007

Dear Mr. Colomb:

On May 3,2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atthe James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The enclosed inspection reportdocuments the inspection results, which were discussed on May 3, 2012, with you and othermembers 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.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 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,de*,^^"f?.Lawrence T. Doerflein. ChiefEngineering Branch 2Division of Reactor Safety Mr. Michael ColombSite Vice PresidentEntergy Nuclear NortheastJames A. FitzPatrick Nuclear Power PlantP. O. Box 110Lycoming, NY 13093

SUBJECT: JAMES A. FITZPATRICK NUCLEAR POWER PLANT. NRC EVALUATION OFCHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANTMODIFICATIONS TEAM INSPECTION REPORT O5OOO333/2012007

Dear Mr. Colomb:

On May 3,2012, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atthe James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The enclosed inspection reportdocuments the inspection results, which were discussed on May 3, 2012, with you and othermembers 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.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 theNRC's document system, Agencywide Documents Access and Management System (ADAI\4S).ADAMS is accessible from the NRC Web site at http:/iwww.nrc.gov/reading-rm/adams.html (thePublic Electronic Reading Room).

Sincerely,/RNLawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyDOCUMENT NAME: GlDRS\Engineering Branch 2\Schoppy\Fitz Mods Report 2012007.docxADAMS ACCESSION NUMBER: M112166A406V suNstReviewg Non-Sensitivetr SensitiveVnPublicly AvailableNon-Publicly AvailableOFFICERI/DRSRI/DRPRI/DRSNAMEJSchoppyMGrayLDoerfleinDATE51301126t1"U126114112RECORDOFFICIAL M. ColombDocket No. 50-333License No. DPR-59

Enclosure:

I nspection Report 05000333/2012007M

Attachment:

Supplemental Informationcc w/encl: Distribution via ListServ M. ColombDistribution Mencl: (VlA E-MAIL)W. Dean, RAD. Lew, DRA(RIoRAMATL RESOURCE)(RIORAMAlL RESOURCE)D. Roberts, DRP (RIDRPMAIL RESOURCE)J. Clifford, DRP (RlDRPMail RESOURCE)C. Miller. DRS(RI DRSMATL RESOURCE)P. Wilson, DRS (RIDRSMAIL RESOURCE)M. McCoppin, Rl OEDOM. Gray, DRPB. Bickett. DRPS. McCarver, DRPM. Jennerich, DRPE. Knutson, DRP, SRIB. Sienel, DRP, RlK. Kolek, Resident AARidsN rrPMFitzPatrick ResourceRidsNrrDorlLpll -1 ResourceROPreports ResourceD. Bearde, DRSL. Doerflein, DRSJ. Schoppy, DRS U.S. NUCLEAR REGULATORY COMMISSIONREGION IDocket No.: 50-333License No.: DPR-59Report No.: 05000333/2012007Licensee: Entergy Nuclear Northeast (Entergy)Facility: James A. FitzPatrick Nuclear Power PlantLocation: Scriba, New Yorklnspection Period: April 16 through May 3,2012Inspectors: J. Schoppy, Senior Reactor Inspector, Division of Reactor Safety (DRS),Team LeaderR. Fuhrmeister, Senior Reactor Inspector, DRSD. Kern, Senior Reactor lnspector, DRSApproved By: Lawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor Safety

SUMMARY OF FINDINGS

lR 0500033312012007i 411612012-51312012; James A. FitzPatrick Nuclear Power Plant(FitzPatrick); Engineering Specialist 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.

REACTOR SAFETY

Gornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R17 Evaluations of Chanqes, Tests, or Experiments and Permanent Plant Modifications(tP 71111.17).1 Evaluations of Chanoes. Tests, or Experiments (20 samples)a. Inspection ScopeThe team reviewed a sample of twenty 10 CFR 50.59 screenings for which Entergy hadconcluded that no safety evaluation was required. The team performed these reviews toassess whether Entergy's threshold for performing safety evaluations was consistentwith 10 CFR 50.59. The sample included design changes, calculations, and procedurechanges and were selected based on the safety significance, risk significance, andcomplexity of the change to the facility. The team reviewed the screenings to determinewhether the changes to the facility or procedures, as described in the Updated FinalSafety Analysis Report (UFSAR), had been adequately reviewed in accordance with10 CFR 50.59 requirements. The team interviewed plant staff and reviewed supportinginformation including calculations, analyses, design change documentation, procedures,the UFSAR, the Technical Specifications (TSs), and plant drawings to assess theadequacy of the screenings. The team compared the screenings and supportingdocuments to the guidance and methods provided in Nuclear Energy lnstitute (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide1

.2.33 support would reduce the stresses for the design basis loading conditions so that theywould be within the USAS 831.1.0 design code allowable stress limits. Entergyperformed an operability determination for CR-JAF-2008-01366, and determined thepiping remained operable but degraded. NRC previously reviewed the operabilitydetermination and documented the results in inspection report 05000333/2008003.Entergy implemented the modification by adding three new pipe supports (PFSK-7082,PFSK-7083 and PFSK-7084) and modifying a fourth existing pipe support (BFSK-771) tothe affected SBGT system piping.The team reviewed the modification to verify that the design basis, licensing basis andstructural integrity of the SBGT piping and supports had not been degraded by themodification. The team interviewed design engineers, and reviewed drawings, pipestress calculations, anchor and support installation procedures, and associatedmaintenance work orders to verify that the SBGT piping support modifications wereappropriately implemented and that the SBGT piping was maintained in accordance withdesign assumptions. The team also performed several walkdowns of the accessibleportions of the modification, and included performing independent measurements, toensure that the system configuration was in accordance with design instructions. Theteam also reviewed corrective action CRs to determine if there were reliability orperformance issues that may have resulted from the modification. Additionally, the teamreviewed the 10 CFR 50.59 screen and engineering evaluation associated with thismodification. The documents reviewed are listed in the attachment.FindinqsNo findings were identified.Electrical Transient Analvzer Prog ram Software VerificationInspection ScopeThe team reviewed a modification (EC 6527) which provided the engineering evaluationfor implementing the Electrical Transient Analyzer Program (ETAP) software program,and included ensuring it met the LevelA requirements of Entergy procedure EN-lT-104,Software Quality Assurance Program. The software program provides a method forelectrical engineers to design and perform studies of electrical power systems. Entergyuses this software to ensure safety-related structures, systems, and components (SSCs)meet their intended design basis functions as defined in FitzPatrick's licensing and designbases documents.The team reviewed the modification to determine if it affected the design or licensingbases or impacted any plant system or component. The team reviewed the designverification checklist, impact screening forms, process applicability checklist,10 CFR 50.59 screening, and engineering evaluation associated with the software toensure that Entergy adequately reviewed the modification. The team reviewed thesoftware's verification and validation documentation provided to Entergy by a qualifiedsupplier to ensure that it was properly verified and validated under the supplier'sEnclosure

410 FR Part 50, Appendix B, quality assurance program. The documents reviewed arelisted in the attachment.b. FindinqsNo findings were identified..2.4 Update Remaininq Service Life for Various Calculations Due to Wall Thinninq lssuesa. Inspection ScopeThe team reviewed a design calculation (EC 11238) which determined the estimatedremaining service life (RSL) for piping in several risk important systems (emergencyservice water, residual heat removal (RHR) service water, RHR, reactor buildingventilation, control room and relay room ventilation, and the torus). Flow acceleratedcorrosion (FAC) is the degradation and consequentialwallthinning of piping orcomponents due to a dissolution phenomenon similar to erosion. Entergy procedure EN-DC-315, FAC Program, established the programmatic criteria and methodology for theinspection, evaluation, and disposition of piping susceptible to degradation due to FAC.Based on ultrasonic test (UT) measurements performed from 2004 to 2008, EC 11238documented the RSL evaluation of 30 pipe segments which had experienced FACdegradation.The team reviewed Entergy's methodology used to determine the monitored susceptiblepiping, measurement techniques, degradation progression, RSL, and Entergy'srecommended corrective actions to ensure piping design code requirements weremaintained. The team selected 11 of the 30 specific pipe wall locations for detailedreview, based on risk insights and estimated RSL of less than 15 years. The teamreviewed the associated piping code design requirements, UT measurement reports, andRSL calculations. Additionally, the inspectors performed selected plant walkdowns andinterviewed engineers to verify that Entergy's monitoring program was effectivelyimplemented, evaluations of RSL were technically sound, safety margins weremaintained, and corrective actions implemented in a timely manner to maintain pipingdesign and associated system operability. Additionally, the team reviewed the10 CFR 50.59 screen associated with this calculation. The documents reviewed are listedin the attachment.b. FindinqsNo findings were identified..2.5 Reactor Recirculation Motor-Generator Set Scoop Tube Positioner Replacementsa. Inspection ScopeThe team reviewed a modification (EC 15323) that replaced the reactor waterrecirculation (RWR) system motor-generator (MG) set scoop tube positioners (02-184ACT-1A & 1B). The function of the RWR MG set scoop tube positioners is to allow theEnclosure b.a..2.65licensed operator to establish precise control of reactor power by setting RWR flow. Theprevious RWR MG set scoop tube positioners were obsolete and were not supported bythe original manufacturer. Entergy implemented the A and B train MG set scoop tubepositioner replacements under child ECs 17564 and 17565, respectively. Entergy useschild ECs to track installation, testing, return to service, and update of configurationdocuments for each separate train when the return to service is completed at differenttimes.The team reviewed EC 15323 and associated EC packages (EC 17564 and EC 17565) toverify that the design basis, licensing basis, and performance capability of the RWRsystem had not been degraded by the modification. The team reviewed the associatedengineering evaluation, engineering change notices, the fire protection programevaluation, the PMTP, work order instructions, vendor manuals for the new equipment,the impact screening form, and the 10 CFR 50.59 screening to verify that Entergyappropriately evaluated the change and developed appropriate installation instructions.The team verified that Entergy identified and appropriately addressed potential effects onthe electrical distribution system. The team discussed aspects of the design and testingwith the engineer responsible for the modification. The team also reviewed EC 26765,Install Digital RWR Recirc Flow Control System, to ensure that Entergy properlyaddressed interfaces between the control system and the controlled component. Theteam reviewed associated cable pull tickets and raceway installation tickets to verify thatEntergy maintained adequate separation of divisional electrical cabling and specifiedappropriate maximum pulling tensions and minimum bend radii. The team performedseveral walkdowns of the RWR MG set scoop tube positioners, accessible electricalraceways, and local indicating panels to ensure that the modification was installed inaccordance with design instructions and to independently assess Entergy's configurationcontrol and the material condition of the RWR MG set room. The team also reviewedcorrective action CRs to verify proper RWR system operation and to determine if therewere reliability or performance issues that may have resulted from the modification. Thedocuments reviewed are listed in the attachment.FindinqsNo findings were identified.Residual Heat Removal Service Water Strainer Coatinq ModificationInspection ScopeSince 2000, Entergy monitored the RHR service water (RHRSW) system strainers forwall thinning through periodic UT examinations. Based on internal inspections of thestrainer housings, Entergy identified that no internal coating remained for thesecomponents. Entergy determined that internal wall thinning of the RHRSW strainersoccurred at an accelerated rate in localized areas of the strainer basket housings. Inresponse, Entergy developed EC 2014 as a mitigating strategy to extend the life of theRHRSW strainers until the strainers could be replaced. The purpose of EC 2014 was toidentify an acceptable coating for the internal surfaces of the RHRSW strainers to preventany further erosion. However, prior to installing the new epoxy coating on the internals ofEnclosure b.a..2.76the old RHRSW strainers, Entergy developed and approved EC 32222 to clarify some ofthe precautions noted in EC 2014 to allow coating application in a shop setting, duringon-line maintenance windows, as part of a large scale replacement of the strainerhousing.The team reviewed EC 2014 and EC 32222 to verify that the design basis, licensingbasis, and performance capability of the RHRSW strainers and the RHRSW system hadnot been degraded by the modification. The team reviewed calculations, engineeringevaluations, strainer specifications, and epory product specifications to verify that theapplied epoxy and modified strainer housings would not adversely impact important tosafety SSCs during normal operation or under design basis conditions. The teamreviewed the associated post-modification test (PMT) results, system health andwalkdowns reports, and corrective action CRs to verify proper strainer operation and todetermine if there were reliability or performance issues that may have resulted from themodification. The team reviewed the associated work order documentation for the A2and 81 RHRSW strainers and performed several walkdowns of the RHRSW strainers toensure that the modification was implemented in accordance with design instructions andto independently assess strainer integrity, Entergy's configuration control, and thematerial condition of the safety-related service water pump rooms. Additionally, the teamreviewed the 10 CFR 50.59 screens and engineering evaluations associated with EC2014 and EC 32222. The documents reviewed are listed in the attachment.FindinqsNo findings were identified.Static Head Correction for Reactor Core lsolation Coolinq and Hiqh Pressure CoolantIniection Suction Pressure InstrumentsInspection ScopeThe team reviewed revised calculations and an associated modification (EC 30006) forthe high pressure coolant injection (HPCI) system and reactor core isolation cooling(RCIC) system pump suction pressure instruments. The pump suction pressureinstruments provide indication, high pressure alarm function, and a turbine trip if suctionpressure falls below the respective pump net positive suction head (NPSH) requirements.Engineers identified that the RCIC instrumentation was not head corrected to address theelevation of the suction pressure transmitter being located 13.625 feet above the pumpsuction elevation. Additionally, the HPCI low suction pressure turbine trip and highpressure alarm functions were not pressure compensated for transmitter elevation (i.e.,static head). Entergy revised the calculations to demonstrate that the corrected RCICand HPCI high pressure alarm setpoints remained below the respective system reliefvalve lift setpoints and that the corrected low suction pressure trip setpoint continued toensure adequate NPSH for the HPCI and RCIC pumps.The team reviewed the modification and revised calculations to verify that the designbasis, licensing basis, and performance capability of the HPCI and RCIC pump suctionpressure indication, overpressure protection, and low pressure protection had not beenEnclosure 7degraded by the modification. The team independently verified the revised RCIC pumpsuction pressure transmitter calibration values were correct and were properly translatedinto maintenance procedures. The team also reviewed the completed PMT and revisionsidentified for the control room simulator. Additionally, the team reviewed the10 CFR 50.59 screen and engineering evaluation associated with this modification. Thedocuments reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.8 Replacement of Reserve Station Service Transformers (71T-2 and 71T-3)a. Inspection ScopeEntergy developed modification EC 12703 to replace the two existing reserve stationservice transformers (RSST) with new transformers of similar impedance. The function ofthe RSSTs is to provide a means of supplying power from the offsite 115 kV power grid(1 15 kV lines 3 and 4) via the 1 15 kV switchyard and stepping it down to the 4160V levelrequired by the plant alternating current (AC) distribution system. The RSSTs provide thepower required for plant start-up and shutdown. These transformers also provide offsitepower to the engineered safeguards equipment for safe shutdown of the plant in theevent of an abnormal or accident condition. The new transformers have a highercapacity rating and have an automatic on-load tap changer (OLTC) capability. Entergyhad planned to install the new RSSTs during Phase 1 of the modification in September2Q10: however, switchyard issues caused Entergy to delay the planned installation untilFall2012. In preparation for the planned installation in September 2010, Entergy hadpulled several hundred feet of associated cable into the East and West safety-relatedcable tunnels prior to RFO19 in 2010.The team reviewed modification EC 12703 to verify that the design basis, licensing basis,and performance capability of the AC distribution system had not been degraded by thepartial implementation of the modification. The team discussed the planned modificationand partially installed portions with design and fire protection program engineers to verifythe design assumptions and program requirements. The team conducted severalwalkdowns and visual inspections of the East and West cable tunnels to assess theinstalled configuration, matedal condition, and potential adverse impact on safety-relatedSSCs in the area. Specifically, the team independently assessed the condition,placement, and storage of the de-energized non safety-related RSST cables for potentialadverse impact on the fire protection system, fire penetration seals, tunnel ventilationsystem, and safety-related cable tray electrical separation. The team also reviewedcorrective action CRs to determine if there were reliability or performance issues that mayhave resulted from the partial implementation of the modification. Additionally, the teamreviewed the 10 CFR 50.59 screen and engineering evaluation associated with thismodification. The documents reviewed are listed in the attachment.Enclosure b.8FindinssNo findings were identified.Screen Wash Booster Pump Discharge Line lsolation Valve Bvpass Modificationlnspection ScopeTwo screen wash booster pumps, arranged in parallel, take suction from the normalservice water (NSW) pump discharge manifold and supply pressurized water to thetraveling water screen wash nozzles. The NSW screen wash booster pump dischargeheader isolation valve, 46MOV-1 1 1, is designed to automatically open when one of thescreen wash booster pumps starts. A FitzPatrick single point failure vulnerability reviewfor the non-safety related circulating water system (LO-WTJAF-2006-1 CA 313)concluded that if valve 46MOV-111 failed to open due to mechanical problems, thetraveling water screens would not start due to low discharge pressure (no flow) atpressure switch 46PS-124, which is located directly downstream of valve 46MOV-111. Inaddition, as 46MOV-111 was the common isolation valve for the discharge line from bothSW screen wash booster pumps, its failure to open would result in both pumps beinginoperable. Entergy developed and implemented EC 14172 to mitigate this single failurepotential by installing a bypass line around 46MOV-111 and included a bypass linemanual isolation valve (46SWS-49). In addition, the modification installed isolation valveson both sides of 46MOV-1 1 1 to enable periodic maintenance on 46MOV-1 1 1 during plantoperation with the bypass line in service.The team reviewed the modification (EC 14172) to verify that the design basis, licensingbasis and performance capability of the screen wash booster pumps and traveling waterscreens had not been degraded by the modification. The team reviewed the associatedwork order instructions and documentation to verify that the modification wasimplemented as designed. The team reviewed related drawings and operatingprocedures to ensure that they were properly updated. The team reviewed theassociated PMTP and PMT results to ensure that Entergy specified appropriate tests andacceptance criteria and that the documented results confirmed satisfactory performance.The team also interviewed plant operators and reviewed corrective action CRs to verifyproper screen wash system operation and to determine if there were reliability orperformance issues that may have resulted from the modification. The team performedseveral walkdowns of the modified screen wash piping and intake area SSCs to ensurethat the modification was installed in accordance with design instructions and toindependently assess Entergy's configuration control and the material condition of theintake area. Additionally, the team reviewed the 10 CFR 50.59 screen and engineeringevaluation associated with this modification. The documents reviewed are listed in theattachment.FindinosNo findings were identified..2.9Enclosure 9.2.10 Service Water Pump Suction Bell Modification: Replace BowlAssemblv with New DesiqnSuction Bell Bearinqa. lnspection ScopeThe team reviewed a modification (EC5000018761) which installed a new design pumpsuction bell and lower shaft on each of the three NSW pumps. The previous NSW pumpsexperienced high vibration and bearing wear. The NSW pumps provide a heat sink forthe reactor building and turbine building closed cooling water systems. Entergy installedthe modified suction bell, with a bearing housing to capture the end of the shaft, to reducebearing wear and pump vibration. The manufacturer estimated that the modified pumpwould have a slightly reduced operating point (flow versus pump total driving head) andslightly increased electrical power demand. Engineers determined that the reduction inpump head and flow and increased power consumption were acceptable because theyrepresented only a small decrease in the existing margin that the pumps had abovesystem operating requirements. Entergy implemented the modification on the three NSWpumps in 2008, 2009, and 2010, respectively.The team reviewed the modification to verify that the design basis, licensing basis andfunctional capability of the NSW pumps had not been degraded by the modification. Theteam interviewed design engineers and reviewed calculations, evaluations, vendor andnameplate data, PMT results, and associated maintenance work orders to verify thatEntergy properly implemented the NSW pump suction bell replacement modification. Theteam verified that NSW flow and pump power consumption did not appreciably changeand that pump vibration was notably reduced. The team also walked down portions ofthe NSW system to observe post-modification pump performance. The team reviewedcorrective action CRs to ensure that Entergy had appropriately addressed any NSWpump modification, PMT, or performance issue. Additionally, the team reviewed the10 CFR 50.59 screen and engineering evaluation associated with this modification. Thedocuments reviewed are listed in the attachment.b. FindinssNo findings were identified..2.11 Residual Heat Removal Service Water Pipinq Remaininq Service Life Calculationa. lnspection ScopeThe RHRSW system is designed to provide cooling water to the RHR heat exchangers(HX). The RHRSW system is operated whenever the RHR HXs are required to operatein the shutdown cooling mode, the suppression pool cooling mode, or the containmentspray mode of the RHR system. The safety-related RHRSW system consists of twoindependent and redundant subsystems. In response to September 2008 UT inspectiondata, Entergy revised calculation JAF-CALC-RHR-03085 (EC 20584) to determine thestructural acceptability of thinned sections of RHRSW piping downstream of 10MOV-89A/B in the A and B RHR HX rooms. The calculation also determined the RSL of thepiping based on the measured wall thickness in the bounding thinned area.Enclosure 10The team reviewed EC 20584 to verify that the design basis, licensing basis, andperformance capability of the RHRSW and RHR systems had not been degraded by theengineering change. Specifically, the team reviewed calculations, technical evaluations,FitzPatrick's piping code and piping specifications, and UT examination history datingback to 1995 to verify that Entergy used appropriate and conservative assumptions toensure that the evaluated RHRSW piping would continue to perform its design functionduring normal operation and under design basis conditions for the calculated remainingservice life. The team compared the calculated piping RSL to Entergy's RHRSW pipingreplacement schedule to ensure adequate margin was maintained and that a reasonableassurance of continued operability existed. The team performed a walkdown of theRHRSW piping in the A and B RHR HX rooms and severalwalkdowns of accessibleRHRSW piping outside the RHR HX rooms to independently assess the RHRSW pipingcondition, Entergy's configuration control, and the material condition of the safety-relatedSSCs in these areas. The team reviewed system health and walkdown reports, andcorrective action CRs to ensure that Entergy had appropriately addressed any RHRSWpiping integrity or performance issues. Additionally, the team reviewed the 10 CFR 50.59screen and engineering evaluation associated with EC 20584. The documents reviewedare listed in the attachment.b. FindinqsNo findings were identified.4.

4OA2 ldentification and Resolution of Problems (lP 71152)a. Inspection ScopeThe team reviewed a sample of CRs associated with 10 CFR 50.59 and plantmodification issues to determine whether Entergy was appropriately identifying,characterizing, and correcting problems associated with these areas, and whether theplanned and/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.4OAO Meetinqs. includinq ExitThe team presented the inspection results to Mr. Michael Colomb, Site Vice President,and other members of Entergy's staff at an exit meeting on May 3,2012. The teamreturned the proprietary information reviewed during the inspection and verified that thisreport does not contain proprietary information.Enclosure

A-1ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

@ger, Design Engineering

R. Casella, Senior Engineer, Mechanical/Structural Design

M. Colomb, Site Vice President

G. Foster, Supervisor, Configuration Management

R. Johnson, Nuclear Plant Operator

S. Juravich, Design Engineer

R. Kester, Senior Engineer, EFIN Structural

K. McWeeny, Senior Reactor Operator

J. Pechacek, Manager, Licensing

A. Porch, Design Engineering

D. Ruddy, Design Engineer

D. Stokes, Fire Protection Engineer

A. Storm, System Engineer

B. Sullivan, General Manager, Plant Operations

A. Yost, Senior Engineer, Electrical Design

LIST OF ITEMS

OPENED, CLOSED AND DISGUSSEDNone.

LIST OF DOCUMENTS REVIEWED

10

CFR 50.59 Screened-out EvaluationsAOP-1 4, Earthquake Process Applicability Determination, dated 41 1 51 10AOP-49, Station Blackout Process Applicability Determination, dated 11114111AOP-64, Loss of Intake Water Level, Rev. 8AOP-68, Spent Fuel PoolTrouble, Rev. 5EC 5681, Equivalency Evaluation of Fluke 8060A Digital Multi Meter, Rev. 0EC 17551, Replace 71UPS-1 MG Set with Static lnverter, Rev. 0EC 19659, Perform Voltage Pickup Calculation Available at 01-125MOV-12(OP) Contactor Coilduring the Degraded Bus Voltage Condition (Refer to

CR-JAF-2010-00242), Rev. 0EC 21686, ESWRHRSW Pump Room Temperature Evaluation - Forced Ventilation ProcessApplicability Determination, dated 4l29l10EC 22434,Install Method to Open 46MOV-111, SWS Screenwash Booster Pump DischargeHeader lsolation Valve, Rev. 0EC 24517, Evaluation of Exxon Polyrex EM Grease VS Chevron SRI Grease for Motor BearingApplications Process Applicability Determination, dated 9111 110EC 26085, LineA/alve Kill Sealant Injection at 35RV-1158, Rev. 0Attachment

A-2EC 30962, Temporary Alarm Set Point Change for 20

TIS-5348 Reactor Building Sump 'B'Temperature, Rev. 2EC 35170, EDG Room Flooding Analysis, Rev. 0EOP-5/6, Secondary Containment Control- Radioactive Release Control Flow Chart Revision 8Process Applicability Determination, dated 81511 1MST-071 .20,125 VDC Station Service Battery Test, Rev. 32OP-45A, Backfeeding Normal Station Service Transformers from the 345 kV System, Rev. 14OP-46A, 4160V and 600V NormalAC Power Distribution, Rev. 56ST-9QA, EDG A and C Full Load Test (8 Hour Run) Process Applicability Determination, dated1t18t11ST-40D, Daily Surveillance and Channel Check Process Applicability Determination, dated1218111TSG-9, Primary Containment Venting Without AC Power, Rev. 4Audits and Self-AssessmentsFebruary 2011 50.59 Screen Sample, Rev. 0February 2012 50.59 Screen Sample, Rev. 0JAFLO-2011-009, Engineering Change Installation Snapshot Assessment, performed2t1t11 - 2t24t11JAFLO-2011-110, James A. FitzPatrick Nuclear Power Plant Focused Self-Assessment Report,dated 416112QA-4-201O-JAF-1, Engineering Quality Assurance Audit Report, dated 8l18l10Calculations04-00457,2004 Pre-Outage Pipe WallThinning Evaluation for System, Rev. 005-00037, Evaluation of R16 Outage UT Inspection Data of Torus, Rev. 011-00012, Static Head Correction for RCIC and HPCI Suction Instruments, Rev. 014620-EM-9011-6, Modification Analysis of Line 24"-G28-152A-46 for Restoration to AcceptableStatus, Rev.014620-EM-9011-7 , Pipe Support Inspection Program Support No.

PFSK-7083

MSK-309E1,Rev.014620-EM-9011-8, Pipe Support Inspection Program Support No.

PFSK-7082

MSK-309E1,Rev.014620-EM-9011-9, Pipe Support Inspection Program Support No.

PFSK-7084

MSK-309E1,Rev.0JAF-CALC-10-00017, ESWRHRSW Pump Room Forced Ventilation, Rev. 0JAF-CALC-RHR-2032, Pipe Wall Thinning Evaluation (A RHRSW), Rev. 0JAF-CALC-RHR-2032

DRN 04-0627, Pipe Wall Thinning Evaluation, dated 10114104JAF-CALC-RHR-2032

DRN 06-02707, Pipe Wall Thinning Evaluation, dated 10l10l0oJAF-CALC-RHR-03085, Remaining Service Life of RHRSW Lines for 2000 Inspections, Rev. 1JAF-CALC-RHR-03085, Remaining Service Life for RHRSW Lines 10-16'-WS-151-30A and 16"-WS-1 51-308122'-WS-1 51 -57, Rev. 2JAF-CALC-RHR-04045, Wall Thinning Evaluation for RHRSW Strainer 10S-5A, Rev. 2AMISC-1835, Evaluation of Acceptable Pit Depth forTorus ShellWall, Rev. 1RHR-3056, Minimum Pipe WallThickness Calculation for RHR Line 16" W20-152-58-RSLCalculation Update, Rev. 1SWS-3118, Minimum Piping WallThickness for System 46, Rev. 0SWS-4170, LocalWall Thinning Evaluation for ESW Strainers 46STR-5A and 46STR-58, Rev. 2Attachment

A-3Condition Reports (CR-JAF-)2008-1 3662008-16272008-33502009-23632009-28932010-0588201 0-09362010-09372010-21252010-25522010-38622010-38932010-61 162010-64232010-7783201 0-7809201

1-01332011-04632011-0572201

1-0951201

1-33932011-54202011-54772011-62762012-17162012-20362012-22072012-22092012-22172012-22242012-22252012-2226*2012-2233.2012-2265*2012-2266*2012-2267"2012-2271*2012-22812012-2289*2012-22902012-23052012-23492012-2354*2012-23592012-23742012-23822012-23882012-23932012-24302012-2443"2012-2492*2012-2540*2012-2549*2012-2556*2012-2559*2012-2562.2012-2570*2012-2581*2012-2582** CR written as a result of this inspectionDesion & Licensinq Bases[[::JAF-DBD-10|JAF-DBD-10]], Design Basis Document for the Residual Heat Removal System, Rev. 13JAF-DBD-16, Design Basis Document for the Primary Gontainment lsolation System, Rev. 4JAF-DBD-46, Design Basis Document for the Normal Service Water, Emergency Service Water,and RHR Service Water Systems, Rev. 18JAF-DBD-68, Design Basis Document for the Drywell Ventilation and Cooling System, Rev. 10JAFP-11-0059, JAFNPP to USNRC, Summary of Plant and Independent Spent Fuel StorageInstallation Changes, Tests, and Experiments for 2009 and 2010 as Required by10

CFR 50.59 and 10

CFR 72.48, dated 5112111USAS 831.1.0, Power Piping, 1967Drawinqs6.60-2, Pipeline Strainer Twin Basket Reheater & Emergency Water Supply, Rev. 126.60-5, 16" Pipe Line Strainer Body ltem No. 10S-5A & 58, Rev. 796-H-600-3-1, Feedwater Heater Setting Plan, Rev.

ABFSK-771, Reactor Building System 27 Containment Vent and Purge Lateral Constraint PipeSupport Detail Civil/Structural, Rev. 3FM-20A, Residual Heat Removal System, Rev. 72FM-20B, Residual Heat Removal Service Water System, Rev. 70FM-30A, Cleanup Filter Demineralizer System, Rev. 38FM-33C, Condensate System, Rev. 27FM-358, Feedwater Heater Vents & Drains, Rev. 24FM-46A, Flow Diagram Service Water System, Rev. 91FP-39A, MISC Piping Screenwell, Rev. 30FSAR Figure No. 9.7-1, Service Water System Flow Diagram, Rev. 12lSl-FM-zOB, lSl Drawing (Residual Heat Removal System), Rev. 14MF-133A, Decay Heat Removal, Rev.

IMSK-137H1, RHR Service Water Piping - Reactor Building System 10, Rev. 13MSK 309E1, Standby Gas Treatment, Rev. 5Attachment

A-4PFSK-7082, Containment Vent and Purge Vertical Lateral Restraint Reactor Building System 27,Rev. 1PFSK-7083, Containment Vent and Purge Dead Weight Support Reactor Building System 27,Rev. 1PFSK-7084, Containment Vent and Purge East-West Lateral Restraint Reactor Building System27, Rev. 1Enqineerino Evaluations2008-1366

CA 01, Engineering Input to Support Operability Determination, dated 4l24lOB2009-2363

CA 01, 46MOV-1 1 1 Apparent Cause Evaluation, dated 8l3l0g2010-2615

CA 01, 46MOV-11 1 Failed to Open Apparent Cause Evaluation, dated 6181102011-0463

CA 02, 105-582 RHRSW Strainer Basket Housing Less Than Minimum WallThickness Apparent Cause Evaluation, dated 2125111EC 14042,lssue ETAP Calculations for Short Circuit and Load Flow Analysis, Rev. 0EC 17248, Child EC to Track Installation, Testing, and Return to Service of

EC 17239 for71lNV-3A, Rev.0EC 17564, Child EC to

EC 15323 to lnstall 02-184ACT-1A, Rev. 0EC 17565. Child EC to

EC 15323 to lnstall 02-184ACT-1B. Rev. 0EC 26765, Digital Recirculation Flow Controls, Rev. 0EC 32222, Clarify RHRSW Strainer Coating

EC 2014 for On Line lmplementation, Rev. 0EN-MA-133 Attachment 9.5, Control of Scaffolding Engineering Evaluation, dated 4127111Single Point Failure Vulnerability Review for Circulating Water System - 036 JAFNPP, May 2008Temporary Modification 7060, LineA/alve Kill at 35RV-1158, Rev. 0Maintenance Work Orders00121241

00199019

00276939 5110074000121642

00199470

00276943 5110074100145981

00212935

00284871 5110074200161192

00212936

00284872 5118388200173597

00247538 0028630300173598

00252248 5110005900178345

00256992 51100739Miscellaneous284871-08, 10S-5A2 PMT, dated 312112284871-09, 10S-5A2 FME Closeout, dated 312112284872-09, 105-581 FME Closeout, dated 3123112284872-13, 105-581 PMT, dated 3123112ASME Boiler and Pressure Vessel Code, Section Vlll, 1986 EditionECT

00199019-15, Scoop Tube Positioner A Functional Test, Rev. 0ECT WO #178345-02, Perform Modification Functional Testing for 46MOV-111 Bypass,performed 10l8l10EN-MA-118 Attachment 9.6, Foreign Material Exclusion Component Close-Out Data Sheet,performed 9124110EPRI

TR-106160, Coatings Handbook for Nuclear Power Plants, June 1996Attachment

A-5lS-M-01 Attachment 2, Surface Preparation and Coating Application Record, dated

3112112 and3t13t12JAF-2009-0511 Action 1,

EC 14172 TEAR Response, dated 11117110JAF-2009-3336

CA 01, IST and SW Programs Action Plan, dated 10114109JAF-2010-001, OSRC Meeting Minutes, dated 1122110James A. FitzPatrick Updated Final Safety Analysis ReportMargin Management List, 4th Quarter 2011Material Request

2577706,2640653,

2716004,

2717435,2736906,

2823704,

2871337,2888017. and2890622MP-059.39, Limitorque Motor Operator Model SB/SMB-000 Corrective and OverhaulMaintenance Requirements, performed 9/2811 0MP-059.43, Maintenance of Limitorque HBC Series Operators, performed 9130110NEI 96-07, Guidelines for 10

CFR 50.59 lmplementation, Rev. 1NP-6695, Guidelines for Nuclear Plant Response to an Earthquake, December 1989NRC Regulatory Guide 1.54, Service Level l, ll, and lll Protective Coatings Applied to NuclearPower Plants, Rev. 2NRC Regulatory Guide 1.187, Guidance for lmplementation of 10

CFR 50.59, Changes, Tests,and Experiments, dated November 2000Refuefing Outage 20 FAC Exam Scope, dated 4118112ModificationsEC 2014, Determine a Suitable Coating for Lining the Interior of the RHR Service WaterStrainers to Prevent AdditionalWallThickness Degradation, Rev. 1EC 6527, ETAP Software Program - Evaluate and Document the Validation, Verification andTesting of the Software per the Requirements of EN-lT-104, Rev. 0EC 7540, Add Supports to Containment Vent and Purge Piping, Rev. 0EC 11238, Update Remaining Service Life for Various Calculations Due to Wall Thinning lssues,Rev. 1EC 12703, Replacement of Reserve Station Service TransformersTlT-2 and 71T-3, Rev. 1EC 14172, Perform Design Change to lnstall a Bypass Valve and lsolation Valves for46MOV-111 to Mitigate the Existing Single Point Vulnerability in the System, Rev. 0EC 15323, Replace Recirc MG Set Scoop Tube Positioners, Rev. 1EC 17239, Replace Obsolete LPCI Inverters, Rev. 0EC 20584, Update of

JAF-CALC-RHR-03085 to Calculate RSL for Wall Thinning Based on 2008UT Inspections, Rev. 0EC 30006, Provide Calculation for Static Head Correction for RCIC Suction lnstruments andHPCI Suction Switches, Rev. 0EC

5000018761, Service Water Pump Suction Bell Modification; Replace Bowl Assembly withNew Design with Suction Bell Bearing, Rev. 1Non-Destructive Examinations and Inspection ReportsO4UT22O,10MOV-89A Ultrasonic Thickness Examination Report, performed 1011310406UT01 1, 10MOV-89A and DS Elbow Ultrasonic Thickness Examination Report, performed6t26t0608UT0122,10MOV-898 D/S Piping Ultrasonic Thickness Examination Report, performed9t22t0808UT0138, 10MOV-89A and D/S Elbow Ultrasonic Thickness Examination Report, performed9t24t08Attachment

A-608UT0143, Piping D/S of 10MOV-89A Ultrasonic Thickness Examination Report, performed9t25t0808UT0151, 10MOV-89A and D/S Elbow Ultrasonic Thickness Examination Report, performed9t25t$82009-3336

CA 04, SW NDE Inspection History,

1115198 - 9125108JAF-RPT-11-00009, FAC Inspection Summary for

RFO-19, Rev. 0JAF'RPT-MISC-01884, Engineering Report Summarizing the Inspection Data for the AugmentedPortion of the JAF FAC Inspection Program, Rev. 9James A. FitzPatrick Nuclear Power Plant Annual Service Water System lnspection SummaryReport, dated 3119110UT-95-003, A RHR SW UT Thickness Readings, performed 3120195UT Erosion/Corrosion Examination Report No. 810UT002, dated 6128110UT Erosion/Corrosion Examination Report No. 810UT003, dated 6128110UT Erosion/Corrosion Examination Report No. B10UT004, dated 6128110UT Erosion/Corrosion Examination Report No. B10UT005, dated 6128110UT Examination Report 06UT017, dated 8/8/06UT Examination Report 08UT024, dated 9l10l0aUT Examination Report 08UT025, dated 9l10l08UT NDE Report Log, dated 9125108Normal and Special (Abnormal) Operations ProceduresAOP-14, Earthquake, Rev. 13AOP-49, Station Blackout, Rev. 18AOP-53, Loss of Spent Fuel Storage Pool, Reactor Head Cavity Well, or Dryer SeparatorStorage Pit Water Level, Rev. 9ARP 09-3-1-9, Fuel Pool Cooling and Cleanup Trouble, Rev. 10ARP 09-6-1-1, Screen Wash BSTR

PMP 46P-6A & 68 Stopped & DISCH MOV Open, Rev. 2ARP 09-6-1-2, Screen Wash Press LO, Rev. 4ARP 09-6-1-9, TRVLG WTR Screen Running for > 30 MlN, Rev. 3OP-308, Decay Heat Removal System, Rev. 15OP-42, Service Water System, Rev. 45OP-45A, Backfeeding the Normal Station Service Transformer from the 345kV System, Rev. 14OP-46A, 4160V and 600V NormalAC Power Distribution, Rev. 56ProceduresAP-19.12, Service Water Inspection Program, Rev. 6EN-CS-S-008-MULTl, Pipe Wall Thinning Structural Evaluation Engineering Standard, Rev. 0EN-DC-117, Post Modification Testing and Special Instructions, Rev. 5EN-DC-126, Engineering Calculation Process, Rev. 4EN-DC-173, Leak Repair Evaluations, Rev. 0EN-LI-100, Process Applicability Determination, Rev. 1 1EN-LI-101, 10

CFR 50.59 Evaluation Program, Rev.9EN-MA-118, Foreign Material Exclusion, Rev. 5EOP-5/6, Secondary Containment Control - Radioactive Release Control, Rev. 8IMP-13.1 , RCIC System Pressure Indication Instrument TesUCalibration, Rev. 15IMP-23.3, High Pressure Coolant lnjection (HPCI) System Flow lndication Calibration, Rev. 22lS-S-02, lnstallation and Inspection of Concrete Expansion Anchors, Rev.22lS-S-04, Pipe Support Installation, Rev. 5Attachment

A-7MP-046.03, Twin Basket Strainers, 46STR-5A(B) and 10S-54(B), Rev. 16MP-059.85, Temporary Leak Repair, Rev. 6MST-071 .20,125VDC Station Battery Service Test, Rev. 32MST-071.29, LPCI Charger-lnverter Performance Surveillance Test, Rev. 13ST-9QA, EDG A and C Full Load Test (8 Hour Run), Rev. 9ST-40D, Daily Surveillance and Channel Check, Rev. 108TSG-9, Primary Containment Venting Without AC Power, Rev. 4Risk and Margin ManasementRisk-lnformed Inspection Notebook for James A. FitzPatrick Nuclear Power Plant, Rev. 2.1Svstem Health Reports & TrendinqJAF-2009-3336

CA 05, SW Component RSL Log, dated 11125109Residual Heat Removal & RHR Service Water Walkdown Report, dated

218112 and 3129112RHR & RHRSW System Health Report, Q4-2011Vendor Technical Manuals and SpecificationsA038-0001 ,24VDC Power Supply, Rev. 0APO-31, Specification for Furnishing and Delivery of Normal Service Water Strainers andEmergency Service Water Strainers, dated 8/8/698953-0002, Gutor Model

PDW 3400-400/600-EA lnverter, Rev. 0Carboguard 890N Product Data Specification Sheet, October 2009J998-0001, Jordan Controls lnstruction Manual

JM-0626 AD-g120 Digital Servo Amplifier, Rev. 0J998-0002, Jordan Controls Instruction Manual

SM-5360 Series Rotary Actuator, Rev. 0JAF-SPEC-MISC-00334, James A. FitzPatrick Nuclear Power Plant Piping Specification, Rev. 14N889-001, NUS

GEN 900 Version 2

NUS-AO73GA Operations and Maintenance Manual, Rev. 1PLMSDS N890A1NL, Carboguard 890 N Part A Material Safety Data Sheet, dated 3125109PLMSDS N890B1NL, Carboguard 890 N Part B Material Safety Data Sheet, dated 1117lO7Attachment

A-8

LIST OF ACRONYMS

AC Alternating CurrentADAMS Agency-Wide Documents Access and Management SystemAOP Abnormal Operating ProcedureASME American Society of Mechanical EngineersCR Condition ReportDBD Design Basis DocumentDRS Division of Reactor SafetyEC Engineering ChangeEntergy Entergy Nuclear NortheastEPRI Electric Power Research InstituteESW Emergency Service WaterETAP ElectricalTransient Analyzer ProgramFAC Flow Accelerated CorrosionHPCI High Pressure Coolant InjectionHX Heat ExchangerlP Inspection ProcedureJAF James

A. FitzPatrick Nuclear Power PlantkV KiloVolt

LOCA [[Loss-of-Coolant AccidentLPCI Low Pressure Coolant InjectionMG Motor-GeneratorMOV Motor Operated ValveNEI Nuclear Energy InstituteNPSH Net Positive Suction HeadNRC Nuclear Regulatory CommissionNSW Normal Service WaterOLTC On-Load Tap ChangerPARS Publicly Available RecordsPMT Post-Modification TestPMTP Post-Modification Test PlanRCIC Reactor Core lsolation CoolingRFO Refueling OutageRHR Residual Heat RemovalRHRSW Residual Heat Removal Service WaterRSL Remaining Service LifeRSST Reserve Station Service TransformerRWR Reactor Water RecirculationSBGT Standby Gas TreatmentSSC Structure, System, and ComponentTS Technical SpecificationUFSAR Updated Final Safety Analysis ReportUSAS]]

USA Standard Code for Pressure Piping

UT Ultrasonic TestUV UndervoltageV VoltAttachment