IR 05000354/2011007

IR 05000354-11-007; PSEG Nuclear, LLC (PSEG); 04/18/2011 - 05/06/2011; Hope Creek Generating Station, Engineering Specialist Plant Modifications Inspection

ML111710487
Person / Time
Site:	Hope Creek, 05000345
Issue date:	06/20/2011
From:	Doerflein L T Engineering Region 1 Branch 2
To:	Joyce T P Public Service Electric & Gas Co
References
IR-11-007
Download: ML111710487 (25)

Text

June 20, 2011

Mr. Thomas P. JoycePresident and Chief Nuclear OfficerPSEG Nuclear LLC-N09P.O. Box 236Hancocks Bridge, NJ 08038

SUBJECT: HOPE CREEK GENERATING STATION - NRC EVALUATION OF CHANGES,TESTS, OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAMINSPECTION REPORT NO. 050003541201 1007

Dear Mr. Joyce:

On May 6,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atyour Hope Creek Generating Station. The enclosed inspection report documents the inspectioniesults, which were discussed on May 6, 2011, with Mr. John F. Perry, Site Vice President, andother 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.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 (ADA[4S)ADAMS is accessible from the NRC Web site at http://www.nrc.qov/readinq-rm/adams.html (thePublic Electronic Reading Room).

Sincerely,--t { ,,D/...r^^-r-.^l 1' l*Lawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyDocket No.: 50-354License No.: NPF-57 Mr. Thomas P. JoycePresident and Chief Nuclear OfficerPSEG Nuclear LLC-N09P.O. Box 236Hancocks Bridge, NJ 08038

SUBJECT: HOPE CREEK GENERATING STATION - NRC EVALUATION OF CHANGES,TESTS, OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAMI NSPECTION REPORT 05000354/2U 1 047

Dear Mr. Joyce:

On May 6,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atyour Hope Creek Generating Station. The enclosed inspection report documents the inspectionresults, which were discussed on May 6, 2011, with Mr. John F. Perry, Site Vice President, andother members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.ln conducting the inspection, the team reviewed selected procedures, calculations and records,observed activities, and interviewed station personnel.Based on the results of this inspection, no findings were identified'In accordance with 1O 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.qov/readinq-rm/adams.html (thePublic Electronic Reading Room).

Sincerely,/RNLawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyDocket No.: 50-354License No.: NPF-57SUNSI Review Complete: LTD (Reviewer's Initials)DOCUMENT NAME: GlDivision\DRS\Engineering Branch 2\Balazik\HCmodsreport2--Q1^1007-.docAflordodarimthisdocum6nl.AnomcialAqenryRo@rd,itwillb9r6i6sed-tolhePublic.ToEc.iw"coPy@ADAMS ACC #ML111710487RECORD COPY T. Joyce

Enclosure:

lnspection Report No. 050003541201 1007M

Attachment:

Supplemental lnformationcc dencl: Distribution via ListServ T. JoyceDistribution Mencl: (via E-mail)W. Dean, RAD. Lew, DRAD. Roberts, DRPJ. Clifford, DRPC. Miller, DRSP. Wilson, DRSA. Burritt, DRPL. Cline, DRPA. Turilin, DRPC. Douglas, DRP(RIORAMAIL Resource)(RIORAMAlL Resource)(RlDRPMail Resource)(RIDRPMAIL Resource)(RlDRSMail Resource)(RlDRSMail Resource)B. Smith, DRP, Acting SRIA. Patel, DRP, RlK. McKenzie, Resident OAS. Bush-Goddard, Rl OEDOD. Bearde, DRSRidsNrrPMHopeCreek ResourceRidsNrrDorlLpl 1 -2 ResourceROPreportsResource@n rc. govL. Doerflein, DRSM. Balazik, DRS Docket No.:License No.:Report No.:Licensee:Facility:Location:lnspection Period:lnspectors:Approved By:U.S. NUCLEAR REGULATORY COMMISSIONREGION I50-354NPF-5705000354/2011007PSEG Nuclear, LLC (PSEG)Hope Creek Generating StationHancocks Bridge, NJApril 18 through May 6, 2011M. Balazik, Reactor lnspector, Division of ReactorSafety (DRS), Team LeaderK. Young, Senior Reactor lnspector, DRSJ. Schoppy, Senior Reactor Inspector, DRSLawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyEnclosure

SUMMARY OF FINDINGS

tR 0500035 412011007; 0411 812011 - 051061201 1; Hope Creek Generating Station; EngineeringSpecialist Plant Modifications Inspection.This report covers a two week on-site inspection period of the evaluations of changes, tests, orexperiments and permanent plant modifications. The inspection was conducted by three regionbased engineering inspectors. The NRC's program for overseeing the safe operation ofcommercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"Revision 4, dated December 2006.No findings were identified.Enclosure 1R17

REPORT DETAILS

REACTOR SAFETYCornerstones: lnitiating Events, Mitigating Systems, and Barrier IntegrityEvaluations of Chanoes. Tests. or Experiments and Permanent Plant Modifications(tP 71111.17)Evaluations of Chanoes. Tests. or Experiments (26 samples)lnspection ScopeThe team reviewed three 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 PSEG had been required toobtain NRC approval prior to implementing the changes. The team interviewed plantstaff and reviewed supporting information including calculations, analyses, designchange documentation, procedures, the UFSAR, the Technical Specifications (TS), andplant drawings to assess the adequacy of the safety evaluations. The team comparedthe safety evaluations and supporting documents to the guidance and methods providedin Nuclear Energy Institute (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," asendorsed by NRC Regulatory Guide 1.187, "Guidance for lmplementation of10 CFR 50.59, Changes, Tests, and Experiments," to determine the adequacy of thesafety evaluations.The team also reviewed a sample of twenty-three 10 CFR 50.59 screenings for whichPSEG had concluded that no safety evaluation was required. These reviews wereperformed to assess whether PSEG'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 PSEG 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 PSEG'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.FindinqsNo findings were identified.a.b.Enclosure

.2.2.1 2Permanent Plant Modifications (11 samples)Service Water Strai ner Structural I mprovementslnspection ScopeThe team reviewed a modification (80097547) that implemented structural improvementsto the safety-related service water (SW) strainers. The SW strainers (H1EA -14, B, C, D-F-S09) had a history of strainer element breakthrough events (failure of the meshscreening when in service during heavy loading). These failures occurred when heavygrass accumulated on the strainer, and increased the differential pressure (DP) acrossthe screen. PSEG modified the construction of the SW strainers to improve the ability ofthe screen to withstand these types of grassing events. Specifically, the structuralimprovements included the installation of an additional mid ring and additional tackwelding on the mesh elements.The team reviewed the modification to verify that the design bases, licensing bases andperformance capability of the SW system had not been degraded by the modification.The team interviewed engineering staff and reviewed technical evaluations associatedwith the modification to determine if the SW strainers and SW system would function inaccordance with the design assumptions. Although PSEG implemented the modificationon all four SW strainers, the team focused their review on the 'A' SW strainer. The teamperformed several walkdowns of the SW pump bays and control room instrumentation toindependently assess PSEG's configuration control, the material condition of the SWstrainers, and the relative grass loading on the SW traveling water screens. The teamreviewed the associated post modification test (PMT) results and SW strainer DP trenddata to verify that the strainers functioned as designed following the modification. Inaddition, the team reviewed several post-modification periodic preventive maintenance(PM) inspections of the SW strainer internals to assess the effectiveness of PSEG'smodification and the condition of the strainers. The team also reviewed corrective actionnotifications (NOTF) to determine if there were reliability or performance issues that mayhave resulted from the modification. The documents reviewed are listed in theattachment.FindinqsNo findings were identified.Reactor Buildinq Floor Pluq Hatch Hold Down Anchors ReconfiourationInspection ScopeThe team reviewed a modification (80097553) that revised the anchoring configurationfor the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC)floor plug hatches in the reactor building (above the HPCI and RCIC pump rooms).PSEG implemented the modification in response to an updated high energy line breakanalysis to ensure that the floor plugs continued to function as designed during designbasis accident conditions. ln addition, PSEG upgraded and revised the anchoringEnclosureb.a..2.2

3configuration to address an adverse condition (corroded bolting) associated with the pre-existing hold down anchors.The team reviewed the modification to verify that the design bases, licensing bases andstructural integrity of the HPCI/RCIC floor plugs had not been degraded by themodification. The team reviewed calculations and technical evaluations to verify that thehold down anchors and floor plugs would function in accordance with designassumptions. The team reviewed the associated work order instructions anddocumentation to verify that maintenance implemented the modification as designed.The team also walked down the floor plugs and adjacent pump rooms to verify thatPSEG had adequately implemented the modification, maintained configuration control,and had not impacted the operation of other safety-related structures, systems, andcomponents (SSC) located in the vicinity. The team conducted a plant walkdown of themodification with the respective design engineer to discuss the critical design aspectsand to independently verify that maintenance installed the anchors in accordance withengineering's design assumptions and associated PSEG procedural requirements.Additionally, the team reviewed the 10 CFR 50.59 screen and engineering evaluationassociated with this modification. The documents reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.3 Hiqh Pressure Coolant Iniection Main Pump and Booster Pump Tie Down BoltReplacementa. Inspection ScopeThe team reviewed a modification (80097504) that replaced the HPCI main and boosterpump tie down bolts. The HPCI main and booster pumps were designed to mount usingdowels and ASTM (American Society for Testing and Materials) 4307 mounting bolts.The dowel pins were designed to bear the horizontal anchor shear loads while the four4307 bolts, per pump, would resist the vertical lifting tension loads. However, during anextent-of-condition review (NUCR 70074025), PSEG had previously identified that thedowel pins, shown on vendor drawings and included in the seismic analysis for resistinghorizontal seismic forces, were not installed in the field. PSEG's operability evaluation(DEH08-0135) for this condition had determined that the dowel pins were not requiredproviding the correct material was used for the mounting bolts. The original 4307 boltswere bearing the full horizontal shear loads in addition to vertical lifting tension loads.However, ASTM 4307 material is a low carbon steel bolt and, therefore, subject to apotential torque relaxation over time which is detrimental to its shear carrying ability. lnorder to address this potential longterm torque relaxation issue, PSEG upgraded the4307 bolts to A325 bolts torqued for a friction type connection per American Institute ofSteel Construction (AISC) requirements. The new HPCI pump anchoring system usingthe ASTM 4325 bolts performs the same design function as the ASTM 4307 bolts anddowel pins, and resolved the corrective actions identified in the associated operabilityevaluation.Enclosure b.a..2.44The team reviewed the modification to verify that the design bases, licensing bases andperformance capability of the HPCI system had not been degraded by the tie down boltreplacement. The team reviewed calculations and technical evaluations to verify that thetie down bolts would function in accordance with design assumptions. The teamreviewed the associated work order instructions and documentation to verify thatmaintenance implemented the modification as designed. The team reviewed theassociated PMT results, HPCI pump vibration and DP trend data, and post-modificationHPCI surveillance test results to verify proper operation of the HPCI system. The teamwalked down the HPCI main and booster pumps to verify that PSEG had adequatelyimplemented the modification, maintained configuration control, and had not impactedthe operation of other safety-related SSCs located in the vicinity. The team alsoreviewed corrective action NOTFs to determine if there were reliability or performanceissues that may have resulted from the modification. Additionally, the team reviewed the10 CFR 50.59 screen and engineering evaluation associated with this modification. Thedocuments reviewed are listed in the attachment.FindinqsNo findings were identified.Emerqencv Diesel Generator Fuel lniector Coolino Water Svstem RemovalInspection ScopeThe team reviewed a modification (80083515) that removed the fuel injector coolingwater system from the'B'emergency diesel generator (EDG). The fuel injector coolingsystem was designed to minimize the effect of fuel varnishing on the fuel injector nozzletips. Fuel varnishing is a particular concern for diesel engines that use a heavy fuel thatmust be heated to remain a liquid. The EDG vendor determined that fuel injector coolingwas of marginal value for Hope Creek's diesel engines because they use a light dieselfuel. The vendor determined that the fuel used to run the EDGs does not need to beheated and the flow of fuel alone provided adequate cooling to minimize the effects offuel varnishing. PSEG identified that this marginal valued fuel injector cooling systemwas the source of minor coolant leakage on all four EDGs and implemented themodification on each of the EDGs to improve EDG reliability, reduce cooling waterleakage, and improve EDG availability.The team reviewed the modification to verify that the design bases, licensing bases andperformance capability of the EDG had not been degraded by the modification. Theteam interviewed engineering staff and reviewed technical evaluations associated withthe modification to determine if the EDG and its support systems would function inaccordance with the design assumptions. The team reviewed the associated work orderinstructions and documentation to verify that maintenance implemented the modificationas designed. The team performed several walkdowns of the four EDGs toindependently assess PSEG's configuration control, the material condition of the EDGs,and the integrity of the cooling water systems. In particular, the team directly observedportions of the 'A' and 'B' EDG monthly surveillances on May 2, 2011 , and April 19,2011, respectively, to assess the cooling water system integrity and performance during=nclosure 5EDG operation at rated conditions. The team also took advantage of an opportunity todirectly observe the as-installed condition of 'B' EDG cylinder No. 2 fuel injector coolingplugs and lock wire, a normally inaccessible area, during an operations inspectionactivity that temporarily removed the cylinder head cover on May 5,2011. The teamreviewed the associated PMT results and several subsequent EDG surveillance testresults, including a 24-hour run, to verify that the 'B' EDG and coolant systemsfunctioned as designed following the modification. The team also reviewed correctiveaction NOTFs to determine if there were reliability or performance issues that may haveresulted from the modification. Additionally, the team reviewed the 10 CFR 50.59 screenand engineering evaluation associated with this modification. The documents reviewedare listed in the attachment.b. FindinqsNo findings were identified..2.5 lnstallation of New Vibration Sensors to Monitor Hope Creek Control Room Supplv andControl Equipment Room Supplv Heatinq Ventilation and Air Conditionino Unitsa. Inspection ScopeThe team reviewed a modification (80072712)that mounted vibration sensors on motorand fan bearings, and installed cable from the sensors through the outside wall of the airhandler to a connection box mounted outside the control area heating, ventilation, andair conditioning (HVAC) equipment (Control Room Supply (CRS) units HlGK-1A-VH403and HlGK-18-VH403, and Control Equipment Room Supply (CERS) units HlGK-1A-VH4O7 and Hl GK-18-VH407). The modification was installed to enhance the ability ofthe operations department and maintenance technicians to periodically collect vibrationdata for the control area HVAC equipment without removing it from service. Themodification provided an external test connection for easy interface of measuring andtest equipment during vibration data gathering activities for the CRS and CERS systems.The team conducted the review to ensure that the design bases, licensing bases, andperformance capability of the CRS and CERS HVAC systems had not been adverselyaffected by the modification. The team reviewed PSEG's installation work orders, whichincluded review of the adequacy of the post-modification testing results. The teaminterviewed the engineering staff regarding the design, installation, and testing of thenew vibration sensors and associated wire harnesses to assess the adequacy of themodification. The team walked down the accessible portions of the new equipment todetermine material condition of the system and ensure the vibration sensors and wireharnesses were installed in accordance with design assumptions and instructions. Theteam also confirmed that surveillance tests, operational procedures, and drawings hadbeen appropriately updated to reflect the design change. The team reviewed correctiveaction notifications, control room HVAC system health reports, and completedsurveillance procedures to determine if reliability or performance issues resulted fromthe modification. The 10 CFR 50.59 screening determination associated with thismodification was also reviewed as described in section 1R17.1 of this report. Thedocuments reviewed are listed in the attachment.Enclosure b..2.6a.FindinqsNo findings were identified.Replacement of Control Room Chiller Bearino Oil and Refrioerant DischarqeTemperature Switcheslnspection ScopeThe team reviewed a modification (80093784) that upgraded temperature switches andthe associated temperature sensor on the 'A' control room chiller (H1GJ-1A-K-400). Thenew temperature switches (H1GJ-IGJTISH-9652A0 and HlGJ-IGJTlSH-9552A2) andsensor (H1GJ-1GJTE-9652A13) are associated with measuring and indicating thrustbearing oil temperatures and refrigerant discharge temperature for the control roomchiller. The upgraded components provided the same function as the previouscomponents. The modification was implemented because the original components wereunreliable and caused operational challenges with the control room chiller system.The team conducted the review to ensure that the design bases, licensing bases, andperformance capability of the 'A' control room chiller system had not been adverselyaffected by the modification. The team reviewed PSEG's installation work orders, whichincluded review of the adequacy of the post-modification testing results. The teaminterviewed the engineering staff regarding the design, installation, calibration, andtesting of the new components to assess the adequacy of the modification. The teamwalked down the accessible portions of the installed equipment to determine materialcondition of the system and ensure the temperature switches and sensors were installedin accordance with design assumptions and instructions. The team also confirmed thatsurveillance tests, operational procedures, and drawings had been appropriatelyupdated to reflect the design change. The team reviewed corrective action notifications,control room chiller water system health reports, and completed surveillance proceduresto determine if reliability or performance issues resulted from the modification. The10 CFR 50.59 screening determination associated with this modification was alsoreviewed as described in section 1R17.1 of this report. The documents reviewed arelisted in the attachment.b. FindinqsNo findings were identified..2.7 Replacement of the 'B' Emerqencv Diesel Generator Fuel Oil Dav Tank Level AlarmSwitch and Raise the Alarm Setpointa. lnspection ScooeThe team reviewed a modification (80080423)that replaced the fuel oil day tank (FODT)level alarm switch (l JELSHL-7501B) and raised the high level alarm setpoint from 9.25inches to 8.75 inches (measured from the top of the tank) for the 'B' EDG FODT. TheFODT level alarm switch (H1JE-lJELSHL-7501B) provides high and low level alarmsEnclosure 7individually at the EDG remote control panel and provides a common trouble alarm inthe main control room. The switch also provides a back-up starUstop function for thefuel oil transfer pumps in case the normal switch fails to operate the pumps. Themodification was installed to correct numerous nuisance alarms and frequent switchfailures during testing of the switch for the FODT. The new component and setpointchange was needed to enhance system reliability and eliminate nuisance alarms.The team conducted the review to ensure that the design bases, licensing bases, andthe performance capability of the EDG FODT system had not been adversely affected bythe modification. The team reviewed PSEG's installation work orders, which includedreview of the adequacy of the post-modification testing results. The team interviewedthe engineering staff regarding the design, installation, and testing of the new FODTlevel switch and revised FODT level setpoint to assess the adequacy of the modification.The team walked down the accessible portions of the new equipment to determinematerial condition of the system and ensure the FODT level switch was installed inaccordance with design assumptions and instructions. The team also confirmed thatsurveillance tests, operational procedures, calibration procedures, and drawings hadbeen appropriately updated to reflect the design change. The team reviewed correctiveaction notifications, the EDG system health reports, and completed surveillanceprocedures to determine if reliability or performance issues resulted from themodification. The 10 CFR 50.59 screening determination associated with thismodification was also reviewed as described in section 1R17.1 of this report. Thedocuments reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.8 Revision to Calculation E-1.4. Hope Creek Class 1E 125 and 250 Vdc Svstems: ShortCircuit and Voltaoe Drop Studies. to Address Batterv lnter-cell Resistancea. Inspection ScopeThe team reviewed a change (S0095526) to calculation E-1.4 which added a basis forthe inter-cell resistance values used in surveillance procedures HC.MD-ST.PJ-0002,"250 Volt Quarterly Surveillance," and HC.MD-ST.PK-0002, "125 Volt QuarterlySurveillance." The procedures provide administrative limits of battery inter-cellresistance of less than 50 micro-ohms. This value is conservative when compared tothe Hope Creek's TS requirement of 150 micro-ohms. The administrative limits in thesurveillance procedures allow PSEG to track, trend, and correct the battery inter-cellresistance values prior to them reaching the TS limit and becoming an operabilityconcern. The calculation change did not impact the inter-cell resistance requirements ofthe TS. PSEG performed this change based on review of an issue identified at anothernuclear power plant.The team evaluated the calculation revision to confirm that the 125 and 250 Vdcbatteries' design bases, licensing bases, and performance capability would not beaffected by the change. The team reviewed the calculation and associated analysis toEnclosure 8verify the assumptions used in the calculation were valid. The team reviewedassociated calculations to ensure they had been updated based on the limits analyzed.The team reviewed procedures to ensure battery inter-cell resistance limits were valid,had been implemented, and were conservative when compared to the applicable TSsections. The team interviewed the engineering staff to review the calculationassumptions, applicable surveillance procedures, and calculation methodology to verifytheir adequacy. The team walked down the battery systems to determine materialcondition of the batteries. The team reviewed corrective action notifications, the 125 and250 Vdc batteries system health reports, and completed surveillance procedures todetermine if reliability or performance issues existed. The 10 CFR 50.59 screeningdetermination associated with this revised calculation was also reviewed as described insection 1R17.1 of this report. The documents reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.9 Service Water Strainer Allowable Differential Pressure Analvsisa. Inspection ScopeThe team reviewed a new analysis (80101059) that ensured the SW strainers (HIEA -1A, B, C, D -F-S09) design function was maintained at elevated differential pressures.PSEG replaced the existing strainers with a more robust design due to the highdifferential pressures experienced during heavy grassing events. The analysis provideda stress calculation to determine the limiting differential pressure across the SW strainer.The analysis resulted in a limiting differential pressure of 101.7 pounds per square inchdifferential. PSEG identified within the analysis that cyclic fatigue failure of the strainerswas not modeled; therefore, maintenance plans were created to replace the strainers ata six-year interval based on past data from strainer failures.The team evaluated the analysis to ensure that the licensing bases, design bases, andperformance capability of the SW system had not been adversely affected by theanalysis. The team reviewed operating procedures to ensure they had been properlyupdated to incorporate the results of the analysis. Drawings were reviewed to verifyinputs into the analysis to ensure proper modeling of the strainers. The team discussedthe analysis and system design basis with both design and system engineers to assessthe adequacy and results of the analysis. In addition, the team verified the SW strainerinstrumentation was adequate for the increased differential pressure. The teamreviewed strainer differential pressure trending data and NOTFs to ensure that the SWstrainer design function was maintained and that margin existed to the limitingdifferential pressure. Finally, the team reviewed the strainer maintenance plan to ensurethe strainer replacement is within the six-year interval. The 10 CFR 50.59 screeningdetermination associated with this modification was also reviewed as described insection 1R17.1 of this report. The documents reviewed are listed in the attachment.Enclosure 9b. FindinosNo findings were identified.2j0 Safetv and Turbine Auxiliaries Coolino Svstem Looic Modificationa. lnspection ScopeThe team reviewed a modification (80098513) that added an automatic function to closethe turbine auxiliary cooling return valve (HV-2496) when the associated supply valve(HV- 2522) inadvertently closes due to failure. These valves function to isolate the non-safety related from the safety-related loads. PSEG implemented this modification toprevent a complete loss of the safety and turbine auxiliaries cooling system (STACS)due to a low expansion tank level, which occurred during past failures of HV-2522. Thecomplete loss of STACS results in a reactor trip. The STACS expansion tank low level,which isolates the corresponding loop, is caused by system sluicing from the operatingloop to the standby loop that occurs during a system load swap when HV-2522 closes'The team's review was performed to verify that the design and licensing bases, andperformance capability of the STACS had not been degraded by the modification. Theteam reviewed a technical evaluation and the system operating parameters, such asvalve sequencing, expansion tank levels, system flowrate, and instrumentation setpointsto verify that margin existed between the low expansion tank level isolation during loopswap. in addition post-modification testing was reviewed to verify proper operation ofSTACS. The team reviewed drawings, procedures, and training documents to ensurethat they were properly updated. Also, the team reviewed documentation to ensure theplant simulator was updated with the modification. The team discussed the modificationand design basis witfr design and system engineers to assess the adequacy of themodification. In addition, the 10 CFR 50.59 screening determination associated with thismodification was also reviewed as described in section 1R17.1of this report' Thedocuments reviewed are listed in the attachment.b. FindinqsNo findings were identified.2.11a.Inspection ScopeThe team reviewed a modification (80102144) that revised calculation H-1-ZZ-MDC-1880 to incorporate a closure time reduction in numerous primary containment isolationvalves along with a change in the methodology of radiological transport of main steamisolation valve leakage. ine modification was implemented to support the approval ofLicense AmgndmeniRequest H09-01 for the use of Cobalt-60 lsotope Test Assembliesin the reactor core. Calculation H-l-ZZ-MDC-1880 evaluates the post loss-of-coolantaccident (LOCA) offsite and control room radiological impact. This modification alsoEnclosurelmpact 10incorporated revisions to H-1-ZZ-MDC-1923 and H-1-ZZ-MDC-1927 , due to the changein H-1-ZZ-MDC-1880. Calculation H-1-ZZ-MDC-1923 addresses radiological impact toareas requiring continuous occupancy during a design bases event and H-1-ZZ-MDC-1927 addresses post-LOCA mission doses to various vital areas in the plant.The team conducted the review to assess whether the calculation revision wasconsistent with assumptions in the design and licensing bases. The team reviewed theassociated revision to the calculations to assess their adequacy and results. The teamverified the results and methodology were in accordance with 10 CFR 50.67 andRegulatory Guide 1

4OA2 ldentification and Resolution of Problems (lP 71152)a. lnspection ScopeThe team reviewed a sample of NOTFs associated with 10 CFR 50.59 and plantmodification issues to determine whether PSEG was appropriately identifying,characterizing, and correcting problems associated with these areas, and whether theplanned or completed corrective actions were appropriate. In addition, the teamreviewed NOTFs written on issues identified during the inspection to verify adequateproblem identification and incorporation of the problem into the corrective action system.The NOTFs reviewed are listed in the attachment.b. FindinqsNo findings were identified.4OAO Meetinos. includinq ExitThe team presented the inspection results to Mr. J. Perry, Site Vice President, and othermembers of PSEG's staff at an exit meeting on May 6,2011. The team returned theproprietary information reviewed during the inspection and verified that this report doesnot contain proprietary information.Enclosure

A-1ATTACHMENT

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

PSEG Personnel

O. Arroyo, Systems Engineer

R. Bhai, Design Engineer

A. Bhuta, Design Engineer

W. Bischoff, Systems Engineer

J. Boyer, Mechanical Engineering Design Manager

D. Boyle, Operations Support Manager

D. Bush, Systems Engineer

P. Duca, Regulatory Assurance Engineer

J. Duffy, Design Engineer

M. Fowler, Engineering Design Manager

A. Ghose, Design Engineer

Y. Ghotok, Systems Engineer

K. Knaide, Engineering Director

J. Moss, Design Engineer

J. Lane, Design Engineer

J. Perry, Hope Creek Site Vice President

G. Siefert, Design Engineer

B. Tarr, Design Engineer

K. Torres, Systems Engineer

K. Yeanrood, Regulatory Assurance Engineer

M. Zimmerman, Design Engineer

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSEDNone

LIST OF DOCUMENTS REVIEWED

10

CFR 50.59 EvaluationsSO.SgEvatuation No.

HC-08-172, Hope Creek Unit 1 Cycle 15 Cycle Management Fuel ChangePackage, Rev. 050.59 Evaluation No.

HC-09-051, Hope Creek Unit 1 Cycle 16 Core Reload Design Fuel ChangePackage, Rev. 150.59 Evaluation No.

HC-10-125, Revise Dose Analysis Calculations and Update UFSAR toRemove 120 Second lsolation for Primary Containment lsolation Valves (PCIV), Rev. 010

CFR 50.59 Screened-out EvaluationsHC-08-134, Reclassification of the Station Service Water System Trash Racks, Rev. 0HC-09-0'15, Revise Calculation

BJ-0001 to Correct Strainer Pressure Loss, Rev. 0Attachment

A-2HC-09-087,

HC.OP-IO.ZZ-0008, Shutdown from Outside Control Room, Rev. 29HC-09-089, Control Room Integrated Display System Computer Replacement-Phase 1, Rev. 2HC-09-102, Replace'A' Fuel Oil Day Tank LevelAlarm Switch and Raise the Alarm Setpoint,Rev.0HC-09-139, Addition of Field Flash Supervisory Light in 1A (8, C, D) Emergency DieselGenerator Exciter Panels 1A (8, C, D) C420, Rev. 0HC-10-024, Diesel Fuel Oil Sampling, Rev. 0HC-10-032, Hope Creek

ECG-RAL Technical Basis Table of Contents, Rev. 0HC-10-049, 'D' Emergency Diesel Generator Jacket Water Keepwarm Reduced WattageOperation, Rev.0HC-10-181, High Pressure Coolant Injection System Operation, Rev.0HC-10-188,

HC.OP-SO.BJ-0001, High Pressure Coolant Injection System Operation, Rev. 4HC-10-191 , Provide Temporary Power to Inverters 1CD481 and 1 CD482, Rev. 0HC-10-207, Shutdown from Outside Control Room, Rev. 30HC-1 1-001 , Revise HC TS Basis 314.9.11, Rev. 0HC-11-004,4.16 kV System Operation, Rev. 2HC-11-006, Diesel Fuel Oil Storage and Transfer System, Rev. 0HC-1 1-010,

CC-AA-1 1 , Non-Conformance Use As ls - Interim - 'B' EDG Lube Oil Filters, Rev. 0Modification Packaqes8OO;72712*, lnstall New Vibration Sensors to Monitor Hope Creek Control Room Supply andControl Equipment Room Supply HVAC Units, Rev. 180080423., Replace 'B' Fuel Oil Day Tank Level Alarm Switch lJELSHL-75018 and Raise theAlarm Set Point, Rev.

080083515, EDG Fuel Injector Cooling Water System Removal, Rev. 280093784., Replacement of the 1AK400 Chiller Bearing Oil and Refrigerant DischargeTemperature Switches, Rev. 280095526., Revise Calculation E-1.4 for Battery lnter-cell Resistance, Rev.

080097504, HPCI Main Pump and Booster Pump Tie Down Bolt Replacement, Rev.

080097547, SW Strainer Structural lmprovements, Rev.

080097553, HPCI/RCIC Floor Plug Hatch Hold Down Configuration, Rev. 080098513., STACS

HV-2496 Closure Logic Change, Rev.

080101059*, Station Service Water Strainers Allowable Differential Pressure lncrease, Rev.

090102144, Revise Calculation s H-1-ZZ-MDC-1880/1

92311927 and lssue UFSAR Change HCN10-014, Rev. 0(. designates a Modification and 10

CFR 50.59 screen-out evaluation sample)Calculations. Analvsis. and Evaluations623-0093, Hatch Plug Pressure, Rev. 16H4-4026, HPCI & RCIC to RACS Floor Plug Hold Down Bolt Replacement, Rev.

070024642, Battery 1A-D-411, Broken Thermometer Found in Cell Number 32, dated 101910270062868, Hope Creek 125V DC Battery Inter-cell Connection Resistance, dated 5l2lo770074346, Extent of Condition for HPCI Pump/Motor Feet Hold Down Bolts, dated 1012610770101272,

EG-2496 22 Second Time Delay Basis Technical Evaluation, dated 9/18/09lO1Og842, HPCI Pump Elevated Vibrations Work Group Evaluation, dated 511711070118122, C SSW Strainer D/P Indication Work Group Evaluation, dated 212411180096269, EDG Fuel lnjector Cooling Water System Removal Modification Acceptance Test,Rev.0Attachment

A-380101 132, lnternal Flood Calculation 19-0018 for Room 5339 Administrative Document ChangeRequest, Rev.080101445, Revise Reactor Overfill Event Calculation

BJ-0024 as a Result of EPU AdministrativeDocument Change Request, Rev. 0BJ-0001, NPSH for HPCI System Pump (Suction from Suppression Chamber), Revs. 5 & 6D7.5, Hope Creek Generating Station Environmental Design Criteria, Rev. 21E-1.4, HC Class 1E 125 & 250V DC Systems: Short Circuit and Voltage Drop Studies, Rev. 6E-17D, HCGS 125V DC: Voltage Drop from Distribution Panel to Load, Rev. 5E-4.2, HC Class 1E DC Equipment and Component Voltage Study, Rev. 4E-5.1, HC Class 1E 250V DC Station Battery and Charger Sizing, Rev. 7EA-0003, Station Service Water System Hydraulic Analysis, Rev. 10EA-01, Station Service Water System Hydraulic Model, Rev. 5Fairbanks Morse Seismic Qualification of Removalof Nozzle Coolant Piping, dated 2127107Fairbanks Morse Un-cooled Fuel lnjector Safety Analysis, dated 311107GE-NE-0000-005-3505-R4, PSEG Nuclear LLC Hope Creek Generating Station Extended PowerUprate, February 2005GM-0027, Diesel Generator Area HVAC Analysis, Rev. 1H-1-EA-MDC-4010, Elastic/Plastic Finite Element Analysis of Hope Creek Service Water StrainerElement, Rev.0H-1-ZZ-MDC-1880, Post-LOCA Exclusion Area Boundary, Low-Population Zone, and ControlRoom Doses, Rev.4H-1-ZZ-MDC-1923, Access to Areas Requiring Continuous Occupancy, Rev. 3H-1-ZZ-MDC-1927, Vital Area Mission Doses, Rev. 2HC-06-14'1, EDG Fuel Injector Cooling Water System Removal, dated 12128106HC-09-01, HPCI Main Pump and Booster Pump Tie Bolt Upgrade, dated 1l2ol09HC-09-010, HPCI RCIC Floor Plug Hatch Hold Down Configuration, dated 1126109JE-0013, Volume of Fuel Oil Day Tank and Level Alarm and Control Set Points, Rev. 7,Root Cause Evaluation

70093613, Reactor Scram due to

HV-2522C SACS/TACS lsolationValve, dated 111712009SC-EG-0087-1, SACS Flow to TACS Heat Exchanger, Rev. 4SC-GJ-0092, Nuclear Boiler Vessel Instrument-AK400 Chiller Comp. Refrigerant Discharge andThrust Bearing OilTemperature High, Rev. 2SC-JE-0057, Diesel Fuel Oil Day Tank Level, Rev. 64SC-JE-0059, Diesel Fuel Oil Day Tank Level, Rev' 6WD 431 173, HPCI Main & Booster Pump Replacement 4325 Tie-Down Bolts, Rev. 0Orders/Evaluations6008942870Q24642700648147007035870Q74155Notification Reports200691 922007278920098332200983987008301 37008526070086733700994977009994920203105202246752026669020301 50670107200701082627011812280091 396800975042033673120362723203629172036420380099080801 01 0598010229120364219203692322037223520375590Attachment

203886322039985820404376204082452040986620411255204383442044112320441430204419972044519920451039204527572Q454804204559352045631220456851204580062046089520461 199204650912Q46750420470663204709722047557620482662204831212049125020495163204964592050372420506263.20506316.2050631 7.20506358.20506593.20506594.20506752*20506754"20506861.20506046.20506059.20506195.20506515"20506520.20506553*20506589.20506592.20506672*20506673.20506674*20506675.20506676.205Q6824.20506880.20506880.20506945.20507220.20507299"20507301*20507302.20507304"20507420.20507520"20507623*20506945.20506952.2050702020507037*205072882050734220507433205078512050809820508099205081 84.205082582050839820508576.2050859920508697.20508698.20508729.20508737.20508739.20508751.2050916420509186.20509253.2Q509281*20509284*20509350(* denotes NRC identified during this inspection)Drawinqs01761770, Electrical Schematic, Engine Control, Sh. 4, Rev. 104-0531-0, Separation Criteria Reactor Building Plan, Rev. 4A-0532-0, Separation Criteria Reactor Building Plan, Rev. 44-0541-0, Separation Criteria Aux Building - Control/Diesel, Rev. 6C-0306-0, Project Civil Standards Floor Hatch Covers, Sh. 2, Rev. 4C-0307-0, Project Civil Standards Floor Hatch Covers, Schedule-Reactor Building, Sh. 3, Rev. 5E-0009-1, Single Line Meter & Relay Dia., 125Y DC Sys., Channels A & C, Sh. 1, Rev. 24E-0009-1, Single Line Meter & Relay Dia.,125Y DC Sys., Channels B & D, Sh. 2, Rev' 28E-0009-1, Single Line Meter & Relay Dia., 125V DC Sys., Channels C & D, Sh. 4, Rev' 13E-0011-1, Single Line Meter & Relay Dia., 250V DC Sys., Sh. 1 & 2, Rev. 17 &18E-0490-0, HCGS, Aux. Building, ControlArea Control Equip. Rm. Supply Fans, Sh. 2, Rev. 2E-0490-0, HCGS, Auxiliary Building and Control Area Control Rm. Supply Fans, Sh. 1, Rev. 4l-0351 1, Strainer Element Assembly 28" Mod. 596 Strainer (Special), Rev.

LM-10-1, Service Water, Sh.1, Rev. 52M-1 1-1, Safety Auxiliaries Cooling Reactor Building, Sh. 1, Rev. 29M-30-1 Sh. 1, Diesel Engine Auxiliary Systems Fuel Oil, Rev. 26M-30-1 Sh. 2, Diesel Engine Auxiliary Systems lntercooler and lnjector Cooling, Jacket Water,Crankcase Vacuum, Air Intake, Exhaust, and Vibration Monitoring Systems, Rev. 20PM018Q-0052, Intercooler & Injector Cooling System, Sh. 1, Rev. 18Attachment

A-5PM723Q-0013, 1gFA Electronic Control Diag. (Chiller 1AK400) for Nuclear Plant Duty, Rev. 14VTD

PN1-E41-C001-002, HPCI Pump Assembly, Sh. 1 & 2, Rev. 6 & 1Procedure/Enqineerino Chanqe DocumentsHCN-09-009, UFSAR Change Request, dated 4128109HCN-10-014 UFSAR Change Request, dated 8119110OTSC # 0A, On-The-Spot-Change,

HC.OP-DL.ZZ-0006-F1,

HC-Auxiliary Building Log 6, dated215110OTSC # 64, On-The-Spot-Change,

HC.OP-AR.GJ-0003, Chiller 1AK400 Control Panel 1AC490,dated 215110OTSC # 48A, On-The-SpofChange,

HC.OP-SO.GJ-0001, 1A(B)K400 ControlArea ChilledWater System Operation, dated 2l5l1oProceduresCC-AA-102, Design Input and Configuration Change lmpact Screening, Rev. 17CC-AA-102-1001, Design lnputs and lmpact Screening-lmplementation, Rev. 5CC-AA-103, Configuration Change Controlfor Permanent Physical Plant Changes, Rev. 13CC-M-103-1001, lmplementation of Configuration Changes, Rev. 3CC-AA-104, Document Change Requests, Rev. 9CC-AA-201, Plant Barrier Control Program, Rev. 1CC-AA-309, Controlof Design Analyses, Rev. 9CC-AA-309-1001 , Guidelines for Preparation and Processing Design Analyses, Rev. 5ER-HC-321-1011, Testing of Hope Creek ASME Class 1,2,3 Safety/Relief Valves, Rev. 1HC.CH-SA.ZZ-0011, Diesel Fuel Oil Sampling, Rev. 21HC.IC-FT.JE-0002, Emergency Diesel Generator B Day Tank Level Switches, Rev. 5HC.MD-PM.EA-0001, Service Water Strainer - Clean and Inspect, Rev. 2HC.MD-ST.PJ-0002, 250 Volt Quarterly Surveillance, Rev. 29HC.MD-ST.PK-0002, 125 Volt Quarterly Battery Surveillance, Rev. 37HC.OP-AB.COOL-0001, Station Service Water, Revs. 16 & 17HC.OP-AB.COOL-0002, Safety and Turbine Auxiliaries Cooling System, Rev. 6HC.OP-AB .ZZ-0001, Transient Plant Conditions, Rev. 22HC.OP-AR.KJ-0003, Diesel Generator Remote Engine Control Panel, Rev. 21HC.OP-AR.ZZ-00A1, Station Service Water system Intake C Trouble Alarm, Rev. 20HC.OP-AR.ZZ-0024, Station Service Water System Strainer C Differential Pressure, Rev. 7HC.OP-DL.ZZ-0006-F 1,

HC-Auxiliary Building Log 6, Rev. 7HC.OP-lS.BJ-0001, HPCI Main and Booster Pump Set - OP204 and OP217 - lST, Rev. 53HC.OP-SO.BJ-0001, High Pressure Coolant Injection System Operation, Revs. 40 and 41HC.OP-SO.EG-001, Safety and Turbine Auxiliaries Cooling Water System Operation, Rev. 44HC.OP-SO.GJ-0001, A(B)K400 Control Area Chilled Water System Operation, Rev. 52HC.OP-SO.GJ-0003, Chiller 1AK400 Control Panel 1AC490, Rev.7HC.OP-SO.JE-OO01, Diesel Fuel Oil Storage and Transfer System Operation, Rev. 28HC.OP-SO.KJ-0001, Emergency Diesel Generators Operation, Rev. 59HC-OP-AR.KJ-0003 ANN C2, Lube Oil Keepwarm Temperature Low, Rev. 21LS-AA-104, 50.59 Review Process, Rev. 6LS-AA-104-1000, 50.59 Resource Manual, Rev.5LS-AA-104-1001, 50.59 Review Coversheet Form, Rev. 2LS-AA-104-1002,50.59 Applicability Review Form, Rev. 2LS-AA-104-1003, 50.59 Screening Form, Rev. 1Attachment

A-6LS-AA-104-1004,50.59 Evaluation Form, Rev. 3MA-AA-716-012, Post Maintenance Testing,Rev. 16MA-AA-734-461, Bolt Torquing and Bolting Sequence Guidelines, Rev. 1NC.DE-TS.ZZ-4009, Station Service Water, Rev. 17NC.EP-EP.ZZ-0202, OSC Activation and Operation, Rev. 17NC.EP-EP.ZZ-0304, Operational Support Center Radiation Protection Response, Rev. 1SH.MD-TI.ZZ-002, Grouting Technical Standard, Rev. 2SM-AA-300-1 001, Procurement Activities and Responsibilities, Rev. 7SY-AA-101-109-1003, Security During Hazardous Exterior Conditions, Rev. 3VHC.MD-ST.GK-0002, Control Room Envelope Inleakage Test, Rev. 1Work Orders3010088530130172301 3803930147637301495453016036230162278301622793016228630162586301 6381 3301654243016542630165691301 71 006301736063017382130182857301 851 66301

86520301

87656301 888993019462230201842302031173020352930205051501134915012295350134384501

35061501 3s165501 351 9550135292501 3659650137201501372065013727250137471501

37580501

37593501 37594501376245013762550137882501

38788501 3962650139627501

39628501

39629501 396305013964750139738501

39739501

39840501 3984150139842501 3984350139844501 39856600564946006276260066851600668526007071360072454600758566007752160077563600775646008071 3600808166008419060084191600841 926008419360089428600900566009288660096424600964256009642660096427Vendor Manuals315177, Pottor & Brumfield Auxiliary Relay Equipment Qualification Report, Rev.

3327152, EDG Fuel Oil Level Switch Seismic Qualification Report, 12l2l04431108, Control Building Chillers, Seismic Qualification Report for Thermowell Part #1S260L10316 and Temperature Sensor Part#RTP775731L10X24, Rev.

2431207, Custom Level Switch Assemblies, 41210980083515 R1 ADM024 R0, Update WD PM018Q-0499 Fuel Injector Cooling Deletion,dated 311107VTD

301835, The Drillco Maxi-Bolt, dated 3/18i05\ffD

322960, Service Water/Strainer Element, dated 417111VTD PM076Q-0042, Service Water: Shop Detailfor Strainer Shoe, dated 414104VTD

431743, Strainer Element Assembly 87727-002 Final Element Analysis and TechnicalCalculation, dated June 2009Attachment

A-7Audits and Self-Assessments30201 132, Quarterly Configuration Control Audit, dated 21211130203529, Quarterly Configuration Control Audit, dated 51611170083142, Procedure Adherence for Configuration Control Program Self-Assessment, dated5t1910870083404, Core Reload and Cycle Management Configuration Changes Using SAP

NF-AA-100-1 001 Self-Assessment, dated 9123108Check-ln Self Assessment of the Hope Creek Setpoint Control Program, dated 6/30/087Q114320, Focused Area Self-Assessment, Evaluation of Changes, Tests, or Experiments andPermanent Plant Modifications, Rev. 1Miscellaneous105431-003, l

EGPSV-2409C Safety Relief Valve Report Form, dated 21110610CFR50.65 (aX3) Periodic Assessment for Hope Creek Generating Station,

211109 - 8131l10125V DC (Class 1E) System Report, Q1- 2011 & Q4-2010250V DC (Class 1E) HPCI/RCIC System Report, Q1-2011 & Q4-201030185109, 12M 1A-F-509 SWIS Strainer Inspection, performed 121211030186526, 12M 1B-F-509 SWIS Strainer Inspection, performed 11124110B Emergency Diesel Operational Monitoring Report, dated 2116111BJ-HPCI System Report, Q1-2011Civil Engineering Studies Structural Research Series No. 353, Behavior of Bolts in TeeConnections Subject to Prying Action, September 1969Control Room Chilled Water System Report, Q1- 2011 & Q4-2010Control Room HVAC System Report, Q1- 2011 & Q4-2010DCR-80095038, Hope Creek Unit 1 Cycle 16 Core Reload Design Fuel Change Package, Rev. 1DCR-80095444,Increase the Calibrated Range for SSW Pressure Diff. Transmitters, Rev. 1Diesel Generators Walkdown Report, dated

11122110 &213111Diesel Generators System Report, Q3-2010 & Q4-2010EA-Service Water System Report, Q1-2011ER-HC-310-1009, Maintenance Rule System Function and Risk Significant Guide, Rev. 6Guide to Design Criteria for Bolted and Riveted Joints, John W. Fisher, Fritz EngineeringLaboratory,1974H-1-ZZ-FEE-1803, Separation Barrier Control Aid for Hope Creek, Rev. 0H-2009-045, SimulatorAction Request, dated 5112109HCG.5-0090, GEH Report TRACGO4P DIVOM 10

CFR 50.59 Evaluation Basis, Rev. 0HPCI Pump Vibration Trend (OP-lS.8J.001), dated

617107 - 4113111KJ-Diesel Generators System Report, Q1-2011NEI 96-07, Guidelines for 10

CFR 50.59 lmplementation, Rev. 1NOH04STACSO, Hope Creek Licensed/Non-Licensed Operator Training-Safety and TurbineAuxiliaries Cooling System, Rev. 7NRC Regulatory Guide 1 .183, Alternative Radiological Source Terms for Evaluating DesignBasis Accidents at Nuclear Power Plants, dated July 2000NRC Regulatory Guide 1.187, Guidance for lmplementation of 10

CFR 50.59, Changes, Tests,and Experiments, dated November 2000NRC Regulatory Guide 1.26, Quality Group Classifications and Standards for Water, Steam, andRadioactive Waste Containing Components of Nuclear Power Plants, dated March 2007Attachment

A-8Service Water System Strainer Differential Pressure Trending Data, dated

111111 to

512111 &111109 to 11110Service Water Walkdown Report, dated

11118110 & 2128111SN N611183-01-0015, Receipt Inspection Report, dated 11130104Technical Specifications, Hope Creek Generating Station, Amendment 185UFSAR, Hope Creek Generating Station, Rev. 17Completed Surveillance & Functional Tests6007752l Operation 0065, W-2 Visual Examination, performed 4120109HC.OP-FT.EA-0001, Validating SSWS Flow Through SACS HXs, performed 3/30/1 1HC.OP-;S.BJ-0001, HPCI Main and Booster Pump Set - OP204 and OP217 - Inservice Test,performed

2124109 and 5/4/09HC.OP-;S.EA-0001, A Service Water Pump - AP502 - Inservice Test, performed 10/5/09 and4t9t11HC.OP-ST.KJ-0002, Emergency Diesel Generator 1BG400 Operability Test - Monthly,performed

2116111,

3115111, and 4119111HC.OP-ST.KJ-0015,

EDG 18G400 - 24 Hour Operability Run and Hot Restart Test - performed10/16/08 and 4114110HC.IC-SC.EG-0007, Safety Auxiliaries Cooling Channel

LT-2508D Expansion Tank

BT-205Level, performed 3111 110HC.IC-DC.ZZ-0074, Baily Single and DualAlarms, Type745, Style 1/2, performed 3111110LAT 3427, Control Room Envelope Inleakage Testing at HCGS Final Report, performed12128109Desiqn & Licensinq BasesDestgn Cntena 10855-D7.3, Procedure for Documentation and Criteria of Plant Separation forthe Hope Creek Generating Station, Rev. 4HCG.5-0081, Nuclear Fuels BWR Design and Analysis Design Input, Documentation ofCompliance with NRC SER Conditions and Limitations for the Application of WCAP-iS141-P-A(Formerly

CENP-287), Fuel Assembly Mechanical Design Methodology forBoiling Water Reactors Supplement 1 to

CENP-287, dated 912108NUREG/CR-0660, Enhancement of On-Site EDG Reliability, February 1979NUREG-0800, Standard Review Plan, Rev. 3Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 184 toFacility Operating License No.

NPF-57 Hope Creek Generating Station, dated 10107110Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 163 toFacility Operating License No.

NPF-57 Hope Creek Generating Station, dated 2108106Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 166 toFacility Operating License No.

NPF-57 Hope Creek Generating Station, dated 417106Specification No. 10855 M-076, Design Specification Service Water Self Cleaning Strainers forthe Hope Creek Generating Station, Rev. 9Attachment

ADAMSAISCASTMCERSCFRCRSDCRDPDRSEDGFODTHCGSHPCIHVACIPLOCANEINOTFNRCPARSPMPMTPSEGRCrCSSCSTACSSWTSUFSARA-9

LIST OF ACRONYMS

Agencywide Documents Access and Management SystemAmerican Institute of Steel ConstructionAmerican Society for Testing and MaterialsControl Equipment Room SupplyCode of Federal RegulationsControl Room SupplyDocument Change RequestDifferential PressureDivision of Reactor SafetyEmergency Diesel GeneratorFuel Oil Day TankHope Creek Generating StationHigh Pressure Coolant InjectionHeating, Ventilation and Air ConditioningInspection ProcedureLoss-of-Coolant AccidentNuclear Energy lnstituteNotificationNuclear Regulatory CommissionPublicty Available RecordsPreventive MaintenancePost Maintenance TestPublic Service Enterprise Group Nuclear

LLCR eactor Core lsolation CoolingStructures, Systems, and ComponentsSafety and Turbine Auxiliaries Cooling SystemService WaterTechnical SpecificationsUpdated Final Safety Analysis ReportAttachment