IR 05000443/2012007
| ML12170A099 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/15/2012 |
| From: | Doerflein L T Engineering Region 1 Branch 2 |
| To: | Walsh K NextEra Energy Seabrook |
| References | |
| IR-12-007 | |
| Download: ML12170A099 (24) | |
Text
-ffi , with copies tothe RegionalAdministrator, Region l; the Director, Office of Enforcement, U.S. NuclearRegulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident lnspector atSeabrook Station. 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.qov/readinq-rm/adams.html (thePublic Electronic Reading Room).
Sincerely,v f)*lLLawrence T. Doerflein, Chief IEngineering Branch 2Division of Reactor SafetyDocket No. 50-443License No. NPF-86
Enclosure:
I nspection Report 05000443/2012007
w/Attachment:
Supplemental Informationcc dencl: Distribution via ListServ In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any)will be available electronically for public inspection in theNRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC's document system, Agencywide Documents Access and Management System (ADAMS).ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (thePublic Electronic Reading Room).
Sincerely,/RNLawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyDocket No. 50-443License No. NPF-86
Enclosure:
I nspection Report 05000443/201 2007
w/Attachment:
Supplemental Informationcc w/encl: Distribution via ListServDistribution Mencl: (via e-mail)W. Dean, RA (RIORAMAlL Resource)D. Lew, DRA (RIORAMAIL Resource)J. Clifford, DRP (RlDRPMail Resource)J. Trapp, DRP (RIDRPMAlL Resource)C. Miller, DRS (RlDRSMail Resource)P. Wilson, DRS (RlDRSMail Resource)A. Burritt, DRPL. Cline, DRPA. Turilin, DRPR. Montgomery, DRPW. Raymond, DRP, SRIJ. DeBoer, Acting RlA. Cass, DRP, Resident AAM. McCoppin, Rl, OEDORidsNrrPMSeabrook ResourceRids N rrDorlLpll -2 Resou rceROPreports ResourceD. Bearde. DRSDOCUMENT NAME: GIDRS\Engineering Branch 2\Mangan\Seabrookmodsreport20l200T.docxADAMS ACCESSION NUMBER: M1121704099g SUNSlReviewg Non-sensitivetr SensitivegnPublicly AvailableNon-Publicly AvailableOFFICERI/DRSRI/DRPRI/DRSRI/DRSNAMEKManganABunittWCookLDoerfleinDATE5t31t1261141126t7t126t15t12RECORD COPYSee Previous Concunence U.S. NUCLEAR REGUI.ATORY COMMISSIONREGION IDocket No.: 50-443License No.: NPF-86Report No.: 0500044312012007Licensee: NextEra Energy Seabrook, LCCFacility: Seabrook Station, Unit 1Location: Seabrook. NH 03874lnspection Period: March 12 through March 29,2012 (on-site inspection period)April 2 through May 7,2012 (in-office review) (part time)Inspectors: K. Mangan, Senior Reactor Inspector, Division of Reactor Safety (DRS),Team LeaderL. Scholl, Senior Reactor Inspector, DRSM. Orr, Reactor Inspector, DRSApproved By: Lawrence T. Doerflein, ChiefEngineering Branch 2Division of Reactor SafetyEnclosure SUMMARY OF FINDINGSlR 0500044312012007; 3112112 - 517112; Seabrook Station, Unit 1 ; Evaluations of Changes,Tests, or Experiments and Permanent Plant Modiflcations.This report covers a two week on-site inspection period of the evaluations of changes, tests, orexperiments and permanent plant modifications. Following the on-site inspection, additional in-office review was performed. The inspection was conducted by three region based engineeringinspectors. One Severity Level lV (SL-lV) violation was identified and characterized as a non-cited violation. The significance of most findings is indicated by their color (Green, White,Yellow, Red) using Inspection Manual Chapter (lMC) 0609, "Significance DeterminationProcess" (SDP). Findings for which the SDP does not apply may be Green or be assigned aseverity level after NRC management review. The NRC's program for overseeing the safeoperation of commercial nuclear power reactors is described in NUREG-1649, "ReactorOversight Process," Revision 4, dated December 2006.NRC-ldentified and Self-Revealinq Findinqs. SL-IV. The team identified a Severity Level lV non-cited violation of 10 CFR 50.59 in thatNextEra made changes to an analysis listed in the Technical Specifications (TS) withoutobtaining a license amendment. The team found that prior to replacing incore probedetectors used to determine neutron and gamma flux in the core NextEra added twocorrection factors to the S3FINC code in order to adjust the signals produced by thedetectors. The changes were made under the 10 CFR 50.59 process. The team alsofound that a third correction factor had been applied in 2002 to address a divergencebetween the measured and predicted flux levels. In this case the changes were madewithout using the 10 CFR 50.59 process. The team's review determined that in 1992 thelicensee had evaluated the methodology used to convert the detector signal to a fluxmap via YAEC-1855PA, Seabrook Station Unit 1 Fixed Incore Detector System Analysis.This analysis had been submitted to the NRC as part of License AmendmentRequest 92-14. The NRC had evaluated and approved the analysis in a SafetyEvaluation associated with License Amendment2T. The analysis was then listed inSection 6.8.1 .6.b.10 of the TS. The team determined that the changes impacted theanalysis and assumptions used as the basis for the conclusions reached in the NRCSafety Evaluation. Following identification of the issue, NextEra entered the issue intothe corrective action program, performed an operability assessment, and planned tocorrect the discrepancy between the license and plant configuration.The team determined that the failure to perform an assessment of the changes made tothe plant in 2002 and that the incorrect conclusion reached in the 2010 10 CFR 50.59evaluation constituted a performance deficiency. Because the issue impacted the abilityof the NRC to perform its regulatory function, traditional enforcement was used todisposition the violation. The issue was considered more than minor because thechanges involved a change to the TS, and the NRC review and approvalwas requiredprior to implementing. The team used IMC 0609, Attachment 4, "Phase 1 - InitialScreening and Characterization of Findings,' to evaluate the risk significance of theissue. The team determined the issue adversely impacted the Barrier IntegrityCornerstone and had very low safety significance (Green) per Table 4a in the Phase 1screening because it only potentially impacted the fuel barrier. (Section 1R17.2.7.b)iiEnclosure Licensee ldentified Violationso A Severity Level lV violation that was identified by NextEra was reviewed by the team.Corrective actions taken or planned by NextEra have been entered into NextEra'scorrective action program. The violation and corrective action tracking number aredocumented in Section 4OA7 of this report.iiiEnclosure REPORT DETAILS1. REACTOR SAFEryGornerstones: Initiating Events, Mitigating Systems, and Barrier lntegrity1R17 Evaluations of Chanqes. Tests. or Experiments and Permanent Plant Modifications(tP 71111.17).1 Evaluations of Chanoes. Tests. or Experiments (25 samples)a. 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 NextEra had been required toobtain U.S. Nuclear Regulatory Commission (NRC) approval prior to implementing thechanges. The team interviewed plant staff and reviewed supporting informationincluding calculations, analyses, design change documentation, procedures, theUFSAR, the Technical Specifications (TS), and plant drawings to assess the adequacyof the safety evaluations. The team compared the safety evaluations and supportingdocuments to the guidance and methods provided in Nuclear Energy Institute (NEl) 96-07, "Guidelines for 10 CFR 50.59 Evaluations," as endorsed by NRC Regulatory Guide1 .187 , "Guidance for lmplementation of 10 CFR 50.59, Changes, Tests, andExperiments," to determine the adequacy of the safety evaluations.The team also reviewed a sample of twenty-two 10 CFR 50.59 screenings for whichNextEra had concluded that a safety evaluation was not required to be performed.These reviews were performed to assess whether NextEra's threshold for performingsafety evaluations was consistent with 10 CFR 50.59. The sample included designchanges, calculations, and procedure changes.The team reviewed the safety evaluations that NextEra had performed and approvedduring the time period covered by this inspection (i.e., since the last plant 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.ln addition, the team compared NextEra's administrative procedures used to controlthescreening, preparation, review, and approval of safety evaluations to the guidance in NEI96-07 to determine whether those procedures adequately implemented the requirementsof 10 CFR 50.59. The reviewed safety evaluations and screenings are listed in theattachment.b. FindinqsNo findings were identified.Enclosure
.2.2.12Permanent Plant Modifications (11 samples)Emerqencv Diesel Generator Enqine Hiqh Temperature Protection Circuit ModificationInspection ScopeThe team reviewed modification EC-12723 that was implemented to improve thereliability of the emergency diesel generators (EDG). The EDGs provide power to the4.16 kV electrical buses to operate safety equipment in the event offsite electrical poweris lost during normal operation, operational transients, or design basis accidents. Duringan accident that results in the initiation of a safety injection (Sl) signal, the EDGprotective trips for generator over current, reverse power, loss of field, high lube oiltemperature, and high jacket water temperature are automatically bypassed as requiredby plant technical specifications. However, EDG protective trips associated with highlube oil temperature and high jacket water temperature trips were not bypassed duringemergency starts not associated with a safety injection actuation, e.9., manualemergency start or loss-of-offsite power condition. This design change modified theEDG control circuitry to bypass these two high temperature trips for all emergency startsof the EDGs. All engine protective trips remain active for testing or normal starts of theEDGs.The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the EDGs had not been degraded by the control circuitmodifications. The team interviewed design engineers and reviewed design drawings todetermine if the circuit changes met the design and licensing requirements. Additionally,the team reviewed post-modification testing (PMT) results, and associated maintenancework orders to verify that the changes were appropriately implemented. The team alsoperformed a walk down of the EDGs and their associated control panels to verify thatannunciators added by with the modification were in accordance with the design and toassess the overall material condition of the systems following the modification work.Finally, the team reviewed affected operating procedures, alarm response procedures,and surveillance test procedures to verify they had been appropriately updated to reflectthe post-modification design and operation. The 10 CFR 50.59 screening determinationassociated with this modification was also reviewed as described in Section 1R17.1 ofthis report. Documents reviewed are listed in the attachment.FindinqsNo findings were identified.4.16 kV Bus-3 and Bus-4 628 Time Delav Relav Setpoint ChanqeInspection ScopeThe team reviewed modification EC-2331 that increased the 628 Agastat time delayrelay setpoint from 1.2 to 2 seconds. The relays monitor voltage on electrical buses 3and 4 which provide 4.16 kV power to non-safety equipment that includes thecondensate pumps and the start-up feedwater pump, and electrical buses 5 and 6 whichEnclosurea.b..2.2 b.3provide 4.16 kV power to safety related equipment. During plant operation buses 3and 5 are powered from the 'A' unit auxiliary transformer (UAT), buses 4 and 6 from the'B' UAT. When the UAT source is lost, buses 3 and 5 automatically transfer to the 'A'reserve auxiliary transformer (RAT) supplied from offsite power. Buses 4 and 6 willtransfer to the 'B' RAT. lf the associated RAT is in synchronism with buses 3 and 5 (or 4and 6), a fast bus transfer to the RAT will occur. lf the buses are not in-sync, the fasttransfer is blocked. When residualvoltage on the buses decays to less than 25 percentof rated voltage a delayed automatic transfer occurs. The 628 time delay relayestablishes the permissive circuit that allows the fast or residualvoltage transfers tooccur. However, if the relay 'times out' all automatic bus transfers are blocked and thebuses must be restored by manual operator actions. NextEra determined the 1.2seconds did not provide sufficient time for the bus voltage to decay below 25 percentvoltage because residual voltage on the 5 (6) bus is produced from the coast down ofthe condensate pump and heater drain pump motors following the loss of the UAT. Thissystem response occurred following a plant event in 2008 when both transfers wereblocked. The time delay setpoint was increased by this modification to allow additionaltime for the bus voltage to decrease below the 25 percent voltage setpoint of the relay toallow the transfer to occur.The team reviewed the modification to determine if the design basis, licensing basis, orperformance capability of the electrical system had been degraded by the modification.In particular, the team verified the time delay change for the non-safety bus transferscheme would not adversely impact the transfer of the safety-related buses. The teaminterviewed design engineers, and reviewed design drawings, PMT results, andassociated maintenance work orders to determine the impact the modification had onthe transfer scheme. The team also verified that associated procedures and drawingshad been updated. Finally, the team walked down the affected switchgear to assess thegeneral material condition of the equipment. The 10 CFR 50.59 screening determinationassociated with this modification was also reviewed as described in Section 1R17.1 ofthis report. Documents reviewed are listed in the attachment.FindinosNo findings were identified.Main Steam lsolation Valve Control Module ReplacementInspection ScopeThe team reviewed modification EC-1451 13 that evaluated and approved the acquisitionof replacement main steam isolation valve (MSIV) control modules. The original modelof circuit cards were no longer available and this design change evaluated theacceptability of using replacement cards that had the same logic and functionalcapabilities as the original cards, but were made with new components and circuit boardsurface mount technology. The main steam isolation valve logic cabinets control theoperation of the MSIVs by providing for the opening, closing, and test functions for thevalves. The control modules provide a portion of the control circuitry contained in the.2.3Enclosure 4logic cabinets. The logic cabinets interface with the engineered safeguards featureactuation system to provide MSIV closure when required for steam line isolation.The team reviewed the modification to determine if the design basis, licensing basis, andperformance capability of the MSIV control circuitry could be degraded by themodification. The replacement modules were currently in stock as spares and nonehave yet been installed in the plant. The team reviewed the associated technicalevaluations and factory acceptance test results, and interviewed design engineers toassess whether the modification was consistent with design assumptions. The teamalso confirmed that the components were procured from a vendor that met the qualityassurance program requirements of 10 CFR Part 50, Appendix B, Quality AssuranceCriteria for Nuclear Power Plants and Fuel Reprocessing Plants. The 10 CFR 50.59screening determination associated with this modification was also reviewed asdescribed in section 1R17.1 of this report. Documents reviewed are listed in theattachment.b. FindinqsNo findings were identified..2.4 Steam Dump Svstem Loss-of-Load Interlock Setpoint Chanqea. Inspection ScopeThe team reviewed modification EC-12711 that evaluated and implemented apermanent change to the main turbine loss-of-load interlock bistable (1-FW-PB-506-C)setpoint. The steam dump system is designed to reduce the magnitude of planttransients following turbine load reductions or turbine trips. To prevent undesired steamdump valve opening following small load perturbations, the steam dump control systemcontains an arming circuit interlock that does not allow steam dump valve openingunless the magnitude of the turbine load reductions is equal to or greater than the loss-of-load interlock setpoint. The original loss of load setpoint was set to arm the valves foropening following a 10 percent step load decrease or a sustained ramp load decrease of5 percent per minute. To prevent undesired arming and opening of the steam dumps,this modification evaluated and implemented a setpoint change to arm the steam dumpvalve opening circuit following a step load decrease of greater than 15 percent.The team reviewed the modification to verify that the design and licensing bases of plantsystems had not been degraded by the set-point change. The team determined ifNextEra had adequately evaluated the impact of the setpoint change on other potentiallyaffected plant analyses and setpoints. The team also verified the calibration procedureshad been updated for the revised set-point and reviewed the results of the completedcalibration performance that implemented the change in the plant. Finally, the teamreviewed the TS to verify that limits in the TS had not been adversely impacted by thechange. Additionally, the 10 CFR 50.59 screening determination associated with thismodification was reviewed as described in Section 1R17.1 of this report. Thedocuments reviewed are listed in the attachment.Enclosure 5b. FindinqsNo findings were identified..2.5 Replacement of 1-Sl-V-32 Motora. Inspection ScopeThe team reviewed modification EC-145087 that evaluated and replaced the actuatormotor for the accumulator isolation valve 1-Sl-V-32. The 1-Sl-V-32 motor is a safety-related motor qualified to operate in harsh environments. The motor is located in thecontainment structure and is credited to operate the 1-Sl-V-32 isolation valve for somebeyond design basis events in order to isolate or un-isolate the safety injection tank.The existing motor was susceptible to internal corrosion and had to be removed from thevalve in order to perform an adequate inspection. NextEra replaced the motor with themotor procured during site construction originally intended for use on this valve.The team reviewed the modification to verify that the design and licensing bases ofsystems had not been degraded by the motor change. The team reviewed theassociated technical evaluations to ensure that the motor had similar electricalcharacteristics to the installed motor. The team also reviewed post overhaul acceptancetest results and interviewed design engineers to assess whether the modificationmaintained the original design requirements. The team also confirmed that thecomponents were procured from a vendor that met the quality assurance programrequirements of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for NuclearPower Plants and Fuel Reprocessing Plants, and that the replacement motor had beenstored in an acceptable storage location. Additionally, the team verified that themaintenance facility that overhauled the replacement motor also had a 10 CFR Part 50,Appendix B, quality assurance program. Finally, the team reviewed the postmaintenance testing to determine if the motor and valve would operate as required. The10 CFR 50.59 screening determination associated with this modification was reviewedas described in Section 1R17 .1 of this report. The documents reviewed are listed in theattachment.b. FindinssNo findings were identified..2.6 Service Water Strainer Basket Modificationa. Inspection ScopeThe team reviewed modification EC-145164 that modified the service water (SW)strainer basket. The basket is credited to collect debris in the SW system prior to thedebris impacting safety-related heat exchangers. NextEra determined duringSW header inspections that debris was bypassing the strainer basket and found that thesupport plate that held the basket to the strainer housing was the source of the debrisEnclosure b.6bypass. The modification extended the surface area of the basket structural supportplate to prevent debris from bypassing the strainer.The team reviewed the modification to verify that the design and licensing bases of plantsystems had not been degraded by the basket modification. The team interviewed thesystem engineer and walked down the system to determine the impact the modificationhad on the SW system. The team also reviewed the post modification testing to ensurethe strainer had been tested in accordance with the American Society of MechanicalEngineers (ASME) code requirements for Class 3 piping. Finally, the team discussedthe corrective actions with the system engineer to determine what follow-up actions werebeing taken to verify the modification had corrected the bypass condition. Additionally,the 10 CFR 50.59 screening determination associated with this modification wasreviewed as described in Section 1R17 .1 of this report. The documents reviewed arelisted in the attachment.FindinssNo findings were identified.Fixed Incore Detector Assemblv Replacement Batch 2lnspection ScopeThe team reviewed modification EC-145087281that evaluated and replaced threepermanently installed incore detector assemblies. The fixed incore detectors are used tomonitor the neutron flux in the core during power operations. The detector assemblies,made up of five detectors mounted vertically in the center of a fuel assembly, convert thelocal neutron and gamma flux levels into a proportional electrical signal. The signalsfrom detectors in 290 core locations are sent to the fixed incore detector data acquisitionsystem and subsequently the S3FINC code to develop the neutron flux profile for theentire core. The modification replaced three detector assemblies including thedetectors, wiring, and electrical connectors/pigtails at the sealtable and containmentpenetration boundary. Additionally, the modification required two constants bedeveloped and incorporated into the S3FINC code. The constants are used to adjust theelectrical signal created by the old and new detectors so that the adjusted signal wouldequate to a signalthat would have been produced by the original detectors when theyhad been newly installed.The team reviewed the modification to verify that the design and licensing bases ofsystems had not been degraded by the detector replacement. The team confirmed thatthe components were procured from a vendor that met the quality assurance programrequirements of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for NuclearPower Plants and Fuel Reprocessing Plants. The team also reviewed that postmodification testing of the equipment verified the proper connection of the wiring.Additionally, the team verified the new seal table connections met the requirement of theASME code. The team reviewed the associated technical evaluations that evaluated ifthe changes to the S3FINC code adequately corrected the probe output signal such thatthe output reflected a consistent measure of neutron flux by the probe. The team alsoEnclosure.2.7 7assessed if the flux measurements remained within the measurement error assumed inthe Core Operating Limits Report (COLR). The 10 CFR 50.59 screening determinationassociated with this modification was reviewed as described in Section 1R17.1 of thisreport. This review included the changes made to the S3FINC code by this modificationand changes made to the code in 2002 to determine if the changes constituted a changeto the methodology. The documents reviewed are listed in the attachment.Findinqslntroduction: The team identified a Severity Level lV non-cited violation of 10 CFR 50.59in that NextEra made changes that affected the TS without obtaining a licenseamendment. Specifically, NextEra made changes to the S3FINC code using their10 CFR 50.59 process; however, the team determined the code had been evaluated andapproved by the NRC in an NRC Safety Evaluation and the analysis was listed inSection 6.8.1 .6.b.10 of the TS.Description: The team reviewed a modification that replaced three Platinum incoredetector assemblies (15 detectors) and the associated 10 CFR 50.59 screen. The teamfound that the modification screen referenced a previous 10 CFR 50.59 screen that hadbeen performed in 2010 when two new detectors assemblies had been installed. Theteam found the previous screen stated, "SPD evaluation will use the S3FINC SPD signalevaluation methodology referenced in the FSAR' and that there was no impact on TSs.The team's review of the modification determined that in order to replace the fixed incoredetector assemblies NextEra added two correction factors to the S3FINC code. The firstcorrection factor was the Platinum depletion correction factor which was applied to eachbatch of detectors to account for detector Platinum 195 depletion. The secondcorrection factor, called the gamma correction factor, was applied to the new detectorsto account for differences in surface geometry of the new detector. Additionally, duringthe inspection the team found that a correction factor had also been applied to theCASMO-3 code in 2002. The code is used to compare the predicted flux to themeasured flux. This correction factor, called the neutron conversion factor, was used toaccount for the transmutation of Platinum 195 to Platinum 196 which results from theexposure of Platinum 195 to a neutron flux. NextEra had concluded that thistransmutation caused detectors to have a decreased sensitivity to gamma flux resultingin a divergence between the profiles.The team reviewed analysis YAEC-1855PA, Seabrook Station Unit 1 Fixed IncoreDetector System Analysis, which the licensee had developed "to demonstrate that thefixed incore detector system is comparable in accuracy and functionality to the standardmovable incore detector system and to define the uncertainty for these powerdistribution measurements." The team found that only one correction to the signalfromthe detectors had been discussed in the analysis. This one-time adjustment, called theraw signal adjustment, was used to adjust the original probe signal outputs to a standardprobe signal output in order to account for probe differences resulting from themanufacturing process. Additionally, the team found that the analysis, YAEC-1855P4,had been submitted to the NRC for review and approval as part of License AmendmentRequest (LAR) 92-14,Incore Detector System.Enclosure 8Finally, the team reviewed the NRC Safety Evaluation in License Amendment 27, whichapproved LAR 92-14. The team found the Safety Evaluation (SE) stated, in part; "Anuncertainty analysis was performed on this data which showed uncertainties of4.13 percent for Fdh and 5.12 percent for Fxy and Fq. These uncertainties are specificto the analytical physics methods, CASMO-3/SIMULATE-3, used at Yankee, the incoredata processing code, FINC, and the Platinum fixed incore detectors currently in use atSeabrook." The SE concluded that "Based on the staffs review of YAEC-1855PA, thestaff finds the methods employed to convert the Platinum detector signal to powerdistribution are mathematically accurate and reasonable from an engineeringstandpoint." The Licensee Amendment resulted in the addition of YAEC-1855P4 to TSSection 6.8.1 .6. b.10., and surveillance requirements in TS 314.2, "Power DistributionLimits," were changed to allow the use of the fixed incore probes.The team concluded that the YAEC analysis could not be changed without a licenseamendment because it is listed as an analysis in the TS. Additionally, the NRC hadapproved LAR 92-14 in the License Amendment 27 based on an evaluation of the entiresignal conversion processing from raw signalto power and the subsequent comparisonof measured to predicted signal as described in the analysis. By adding three correctionfactors to the signal processing of the probes, NextEra had changed the YAEC analysisas approved by the NRC and, therefore, potentially invalidated the conclusions reachedin the associated Safety Evaluation. Based on the team's conclusions NextEra enteredthe issue into the corrective action program, performed an operability assessment, andwill evaluate actions to ensure the license and plant configuration are in alignment.Analvsis: The team determined that the failure to perform an assessment of thechanges made to the CASMO-3 code in 2002 and the incorrect assessment performedin the 10 CFR 50.59 evaluation related to the 2010 changes to the incore detectorassemblies was a performance deficiency. Specifically, in both cases a discussion ofthe impact the changes had on the analysis listed in the plant Technical Specificationswas not performed, and the changes potentially invalidated the conclusions reached inthe NRC Safety Evaluation. Because the issue impacted the ability of the NRC toperform its regulatory function, the team evaluated the issue using the traditionalenforcement process. The violation was more than minor because there was areasonable likelihood that the changes requiring the 10 CFR 50.59 evaluation wouldrequire NRC review and approval prior to implementation. The team used IMC 0609,Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," to evaluatethe risk significance of the issue. This finding adversely impacted the Barrier lntegrityCornerstone objective of providing reasonable assurance that physical design barriers(fuel cladding) protect the public from radionuclide releases caused by accidents orevents. Specifically, this finding challenged the design control attribute to ensure thefuel cladding is maintained within the established limits of the core operating limit reportand reload analysis. The issue was determined to have very low safety significance(Green) per Table 4a in the Phase 1 screening because it only potentially impacted thefuel barrier. Traditional Enforcement violations are not screened for cross cuttingaspects.Enforcement: 10 CFR 50.59 (c)(1)states in part that the licensee may make changes inthe facility as described in the FSAR (as updated) without obtaining a licenseEnclosure a.Iamendment pursuant to 10 CFR 50.90 only if. (i) A change to the TSs incorporated inthe license is not required. Contrary to the above in2002 and in 2010 NextEra madechanges that impacted the YAEC-1855PA analysis without obtaining a licenseamendment and a license amendment was required. Specifically, this analysis waslisted in the TS and approved by the NRC and, therefore, a license amendment inaccordance with 10 CFR 50.90 was required. ln accordance with the NRC EnforcementManual Section 7.3, "Enforcement of 10 CFR 50.59 and Related FSAR," and NRCEnforcement Policy Section 6.1, the violation was given a Severity Level lVcharacterization because the issue was evaluated to be of very low safety significance.However, because this violation was determined to be of very low safety significanceand was entered into the corrective action program (AR 1761442), this violation wastreated as a non-cited violation, consistent with Section 2.3.2 of the NRC EnforcementPolicy. (NCV 05000443/2012007-001, Inadequate 10 CFR 50.59 Evaluation)New Non-metallic Liner for Spent Fuel Poollnspection ScopeThe team reviewed modification EC-156597 that replaced the liner installed on thestainless steel plates in the cask loading and fuel transfer canal areas of the spent fuelpool (SFP). The original liner had been installed in 2004 in order to provide a water tightbarrier between the SFP water and the stainless steel plating and associated welds.The non-metallic liner design function is to prevent the leakage of tritiated water fromSFP. Due to indications of degradation, NextEra determined that the liner needed to bereplaced. The new non-metallic liner is made up of a primer coating and an epoxy resinapplied to the stainless steel liner. ln addition to the installation of the liner, themodification established an inspection program to monitor the new liner and the epoxycoupons that were created during the modification.The team reviewed the modification to verify that the design and licensing bases of plantsystems had not been degraded by the replacement of the liner. The team reviewed theinstallation procedure to verify the surface had been properly prepared and the materialhad been installed in accordance with the manufacturers' recommendations. The teamalso reviewed the results of the post installation inspection to determine if deficiencieshad been identified. and to determine if corrective actions had been taken to address thedeficiencies. The team reviewed NextEra's program that was developed to determinethe adequacy of bonding capability of the non-metallic liner to the stainless steel linerprior to installing the product. Additionally, the team reviewed the program formonitoring the liner, which included the testing program for adhesion of the material tocoupons placed in the SFP, developed to verify the integrity of the liner followinginstallation. The team conducted this review to ensure that degraded coating would nottransit to the reactor vessel during refueling operations. Finally, the 10 CFR 50.59screening determination associated with this modification was reviewed as described inSection 1R17.1 of this report. The documents reviewed are listed in the attachment.FindinqsNo findings were identified.Enclosureb.
a.10Ambient Temperature Monitorinq East and West Pipe ChaseInspection ScopeThe team reviewed modification EC 2334 that installed temperature indicators in thevicinity of the feedwater isolation valves (FlV). NextEra installed local temperatureindicators in order to monitor ambient air temperature at the middle level of each pipechase. The modification was performed after it was determined that both nitrogenpressure and the hydraulic fluid flow characteristics for the FlVs could be affected byambient and componenUpiping temperatures. Specifically, NextEra recognized that lowambient temperatures had an adverse influence on the stroke time of the FlVs. Themodification provides data used by operators to monitor FIV operability during the wintermonths.The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the FlVs had not been degraded by the modification. Theteam interviewed engineering staff and reviewed technical evaluations associated withthe modification to determine if the instruments would provide adequate data to supportproper operation of the valves. The team reviewed drawings, procedures, andmaintenance plans to ensure that they were properly updated or developed. The teamalso performed walkdowns of both pipe chases to verify the installation was completedas designed and to assess the material condition of the equipment. Additionally, the10 CFR 50.59 screening determination associated with this modification was reviewedas described in Section 1R17.1 of this report. The documents reviewed are listed in theattachment.b. FindinqsNo findings were identified..2.10 Substitution of Mobil DTE732 for Mobil 797a. Inspection ScopeThe team reviewed modification EC 2520 authorizing the substitution of Mobil DTE 732for Mobil 797 oil in all applications at the Seabrook Station. The equipmenVapplicationsaffected by the change included charging pumps, containment building spray (CBS)pumps, EDG governors, feedwater pumps and turbines, and Sl pumps. The substitutionwas required as a result of Exxon/Mobil discontinuing production of Mobil 797.The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the various systems or components had not been degraded bythe modification. The team reviewed vendor product literature along with technicalevaluations to ensure compatibility of the two products if mixing were to occur in variousquantities. The team interviewed engineering staff to determine if compatibility issueshad arisen since the introduction of the new product. Additionally, the team reviewed asampling of affected system maintenance procedures to verify the reviseddocumentation contained the appropriate information. Finally, the team walked downEnclosure 11the accessible portions of the Sl, CBS and charging systems, and the EDGs to assesstheir material condition. Additionally, the 10 CFR 50.59 screen associated with thismodification was reviewed as described in Section 1 R17 .1 of this report. Thedocuments reviewed are listed in the attachment.b. FindinqsNo findings were identified..2.11 Service Water Pipinq Repair to 1-SW-1814-1-156-24a. Inspection Scopeb.The team reviewed modification EC 145189 that repaired a through-wall leak on SW line1-SW-1814-1-156-24 by full penetration welding of a PMCap on the outside of the pipeover the degraded area. The PMCap is a pre-engineered pipe cap designed to meetASME Section lll Class 3 requirements. The proximity of the repair to piping support No.1804-SG-02 additionally necessitated a modification to that support. The function of theSW system is to transfer the heat loads from various sources in both the primary andsecondary portions of the plant to the ultimate heat sink.The team reviewed the modification to verify that the design basis, licensing basis, andperformance capability of the SW system had not been degraded by the modification.The applicable ASME Boiler and Pressure Vessel Code and interpretations werereviewed to ensure NextEra's justification and design assumptions were acceptable.The team ensured pipe stress calculations were appropriately revised to account for thepermanent removal of the east-west restraint function of piping support No. 1804-SG-02and addressed the revised loading on other pipe supports. The team reviewed theultrasonic test examination data, the PMCap vendor product and installation drawings,work order packages, and photographs of the finished installation. Additionally, the teaminterviewed engineers to confirm that the repair satisfactorily functioned in accordancewith design assumptions. A walkdown of the accessible system piping was performed toassess material condition. The 10 CFR 50.59 screen associated with this modificationwas reviewed as described in Section 1R17 .1 of this report. The documents reviewedare listed in the attachment.FindinosNo findings were identified.OTHER ACTIVITIESfdentification and Resolution of Problems (lP 71152)Inspection ScopeThe team reviewed a sample of condition reports (CR) associated with 10 CFR 50.59and plant modification issues to determine whether NextEra was appropriatelyEnclosure4.4c.A2a.
12identifying, characterizing, and correcting problems associated with these areas, andwhether the planned or completed corective actions were appropriate. In addition, theteam reviewed 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. FindinssNo findings were identified.4OAO Meetinqs. includinq ExitThe team presented the preliminary inspection results to Mr. Paul Freeman, Site VicePresident and other members of NextEra's staff at a meeting on March 29, 2012.Following additional in-office review, which included support from the Office of NuclearReactor Regulation (NRR) staff, the team conducted a final exit meeting on May 7,2012,via teleconference with Mr. Freeman and other members of NextEra's staff to discussresults of the inspection. The team returned proprietary information reviewed during theinspection and verified that this report does not contain proprietary information.4OA7 Licensee-ldentified ViolationsThe following violation of Severity Level lV was identified by the licensee and is aviolation of NRC requirements which meets the criteria of the NRC Enforcement Policyfor being dispositioned as a Non-Cited Violation.o A violation of 10 CFR 50.59 was identified in that NextEra improperly made a changeto the UFSAR using the 10 CFR 50.59 process that required NRC review andapproval prior to implementation. Specifically, in 2010, NextEra changed a graph inthe UFSAR used to determine the reactivity in a fuel bundle based on burn-up sothat the fuel could be safely placed in the spent fuel pool. This change to the UFSARwas made using the 10 CFR 50.59 process. However, because this same graph isin the Technical Specifications, the 10 CFR 50.59 process could not be used and alicensee amendment under 10 CFR 50.90 was required. Traditionalenforcementapplied because the change impeded the regulatory process. The issue is morethan minor because the change was made without NRC review and approval prior toimplementation. The issue was determined to be of very low safety significance(Green) using IMC 0609, Attachment 4, because it was a spent fuel pool issue thatdid not result in a loss of SFP cooling, did not involve a fuel handling error, and didnot result in a loss of SFP inventory. Subsequent to making the change to theUFSAR, NextEra submitted a license amendment request (LAR 11-04) in January2012 to obtain approval for the change. NextEra entered the issue into thecorrective action program (CR 1744734) for evaluation and resolution.Enclosure Licensee PersonnelP. FreemanR. NobleM. CollinsJ. SobotkaM. O'KeefeP. GurneyG. KotkowskiD. YatesR. DeanR. PerryC. CroninH. MentelN. PietrantonioV. BrownT. CarterR. DeanC. MelloT. NagleT. SchulzJ. SweeneyE. TrumpJ. EstevesK. RandallA. MerrillA-1ATTACHMENTSU PPLEMENTAL INFORMATIONKEY POINTS OF CONTACTSite Vice PresidentEngineering DirectorDesign Engineering ManagerSystem Design ManagerLicensing ManagerReactor Engineer SupervisorElectrical Design Engineering SupervisorSystem EngineerPrincipal Engineer l&CCBM IST GroupDesign EngineerDesign EngineerDesign EngineerRegulatory AssuranceDesign EngineerDesign EngineerDesign EngineerDesign EngineerDesign EngineerDesign EngineerFire Protection EngineerDesign EngineerReactor EngineerReactor EngineerITEMS OPENED, CLOSED AND DISCUSSEDOpened and Closed05000443/2012007-001NCV Inadequate 10 CFR 50.59 EvaluationLIST OF DOCUMENTS REVIEWED10 CFR 50.59 Evaluations08-005, Turbine Control System Replacement, Rev. 009-001, Risk Informed lSl of Class 2 Main Steam and Feedwater Break Excursion Region,Rev.010-003, Spent Fuel PoolAdministrative Controls, Rev. 0Attachment A-210 CFR 50.59 Screened-out Evaluations03-212, Boral Blistering Non Conformance Report, Rev. 008-052, Ambient Temperature Monitoring East and West Pipe Chase, Rev. 0*08-495, Substitution of Mobil DTE732 for Mobil 797, Rev. 0*09-227, Replacement of 1-Sl-V-32 Motor, Rev. 0*09-063, Fixed Incore Detector Assembly Replacement Demonstration, Rev. 209-323, Service Water Strainer Basket Modification, Rev. 0*09-336, SW Piping Repair to 1-SW-1814-1-156-24, Rev. 0*10-065, Reconciliation of Vortex lssue for the Condensate Storage Tank, Rev. 010-091, PORV Block Valve Operation, Rev. 010-099, Fixed Incore DetectorAssembly Replacement Batch 2 - OR14, Rev.2*10-121, EDG Engine Temperature Control Upgrade, Rev. 110-136, New Non-Metallic Liner for Spent Fuel Pool, Rev. 0*10-271, Replacement of SCO-ST Relay for Rx Coolant Pump Locked Rotor Protection, Rev. 010-032, Reactor Trip or Safety Injection Emergency Operating Procedure Change, Rev. 01't-133, Service Water Cooling Tower Operation, Rev. 011-243, Service Water Pump 'D' Motor Winding Temperature High, Rev. 011-281, MSO Resolution: EFW Flow ControlValve Circuit Changes, Rev. 011-350, Revise Emergency Feedwater High Flow lsolation Setpoint, Rev. 0(* designates a 10 CFR 50.59 screen-out evaluation sample that was also a modification sample)Modification PackasesEC 12711 (08 MMOD 506), Steam Dump Loss of Load Arming Setpoint Change, DCN 00 & 01*EC 12723 (08 MMOD 518), Emergency Diesel Engine High Temperature Protection CircuitModification, DCNs 00-06"EC 145087, Replacement of 1-Sl-V-32 Motor, Rev. 1*EC 145113, Main Steam lsolation Valve Control Module Replacement, Rev. 0*EC 145164, Service Water Strainer Basket Modification, Rev. 1"EC 145189, SW Piping Repairto 1-SW-1814-1-156-24, Rev. 0*EC 145281, Fixed Incore Detector Assembly Replacement Batch 2 - OR14, Rev. 4*EC 156597, New Non-Metallic Liner for Spent Fuel Pool, Rev. 0*EC 2331 (08 MSE 021),4.16 kV Bus-3 & Bus-4 628 Time Delay Relay Setpoint Change, Rev. 0*EC 2334 (08 MSE 024), Ambient Temperature Monitoring East and West Pipe Chase, Rev. 0*EC 2520 (08 MSE 210), Substitution of Mobil DTE732 for Mobil 797, Rev. 0*(* designates a modification sample that is also a 10 CFR 50.59 screen-out evaluation sample)Calculations. Analvsis. and Evaluations0570-021-003, Lubrication Data Sheets/Thermal and Radiation Properties, Rev. 108-01003, Apparent Cause Evaluation: Feedwater lsolation Valve FW-V-30 Failed Stroke Time,Rev.0CN-CPS-07-30, Seabrook Steam Dump Arming Setpoint Analysis, Rev. 1C-S-1-45560, SW Spools in Tank Farm/PAB Service Water Piping, Rev. 4EC-274551, Revise EFW High-Flow lsolation Setpoint, Rev. 0EE-07-035, Engineering Evaluation - Risk Informed Inservice Inspection of Class 2 Main Steamand Feedwater Break Excursion Region, Rev. 0EE-08-006, Engineering Evaluation for FWIV Ambient Temperature Assessment, Rev. 0Attachment A-3EQ File 225-03-01M, Environmental Lubrication Evaluation, Rev. 0NSS-220-01, Environmental Qualification of Electrical Equipment, Rev. 4PM Technical Basis: 1-EAH-Tl-8600/8601 Run to Failure Basis, dated 911109EC-07-030, Status of Fixed/Movable lncore Detector System, Rev. 1Condition Reports00000903 01744734* 01761442" 08-0103900046104 01748223* 07-00944 08-0119700091229 01748223* 03-02019 08-0120400208369 01748224" 07-07449 08-1360901646426 01748224. 07-1192801681835 01749138* 08-0089401744075* 01750077" 08-00978(* denotes NRC identified during this inspection)Drawinqs118716G52, Sht. 5, Wiring Diagram Power Unit, Rev. 121-MAH-820503, Miscellaneous Air Handling Details, Rev. 151-NHY-31002, Electrical Distribution One Line Diagram, Rev.421-NHY-31009, Shts. 1 and 2,4160V Switchgear Buses 1-3 and 1-4 One Line Diagram, Revs. 14and21-NHY-310102, Sht. G10/b, DG-1A Monitoring & Auxiliary Control Schematic Diagram, Rev. 111-NHY-310102, Sht. G1073d, Diesel Generator 1A AnnunciatorAuxiliary Relays SchematicDiagram, Rev. 101-NHY-310102, Sht. G19/3d, Diesel Generator 1B AnnunciatorAuxiliary Relays SchematicDiagram, Rev. 111-NHY-310102, Sht. G20b, DG-18 Monitoring & Auxiliary Control Schematic Diagram, Rev. 121-NHY-310857, Sht. E93/8a, Emergency Diesel Generator 1-A Start Circuit No. 1 SchematicDiagram, Rev.91-NHY-310857, Shts. E93/8c, e, and f, Emergency Diesel Generator 1-A Start Circuit No. 2Schematic Diagram, Revs. 1 1, 8, and 81-NHY-310857, Shts. E94l8c, f, and e, Emergency Diesel Generator 1-B Monitor CircuitSchematic Diagram, Revs. 9, 8, and 81-NHY-503491, Sht. 2, DG-DieselAir Start Logic Diagram, Rev. 01-NHY503492, DG-Diesel Shut-Down and Emergency Stop Logic Diagram, Rev. 111-NHY-506425, EAH Main Steam and Feedwater Pipe Enclosure Control Loop Diagram, Rev. 61-SW-D20795, Service Water System Nuclear Detail, Rev. 40200917-M-0001, Shts. 1-3, Service Water Pipe 1-SW-1814-156-24 PMCap Shop FabricationDetails, Rev.2200917-M-0002, Sht. 1, Service Water Pipe 1-SW-1814-156-24 PM Cap Installation Details,Rev.09763-F-500221, Main Steam and Feedwater East Pipe Chase Instrument Piping, Rev. 8Dwg. 128417, SW Strainer Baskets, Detail, SW-9-10&11, Rev. 11MSE-08024-2000, Add Temperature Indicators PID MAH-820503, Rev. 0Partial Dwg. 5531-SG-7, Installation DetailforTemperature !ndicators EAH-TI-8600 and 8601,Rev.0PID-1-SW-D20795, Service Water System, Rev. 40SK-1000, Modification to Support 1-1814-SG-02, Rev. 0Attachment A-4SK-EC145164, Service Water Strainer Basket Modification, Rev. 1SK-EC145189-2000, SW System Piping Repair 1-SW-1814-1-156-24, Rev. 0SW-1827-08, Sht. 1, Service Water 1-SW-1827-1, Rev. 13Licensinq DocumentsAmendment No. 27 to Facility Operating License NPF-86, Incore Detector System - LicenseAmendment Request 92-14, dated 12122193Amendment No. 6 to Facility Operating License NPF-86, Seabrook Station, Unit No. 1, dated8t27191Letter from North Atlantic Energy Services to USNRC, "License Amendment Request 92-14,lncore Detector System," dated 11125192Letter from North Atlantic Energy Services to USNRC, "Response to Request for AdditionalInformation: License Amendment Request 92-14," dated 712193Letter from North Atlantic Energy Services to USNRC, "Response to Request for Additionallnformation: License Amendment Request 92-14," dated 11124193Seabrook Station Final Safety Analysis Report, Rev. 14,ASeabrook Technical Specifications, Rev. 9-29YAEC-1855PA, Seabrook Station Unit 1 Fixed Incore Detector System Analysis, dated 10192Procedures5059RM, 50.59 Resource Manual, Rev. 10DG-CP-7SA UA-9558, Panel DG-CP-7S UA-9558 LocalAlarm Responses, Rev. 53DG-CP-76A UA-9568, Panel DG-CP-76 UA-9568 LocalAlarm Responses, Rev. 55E-0, Reactor Trip or Safety lnjection, Rev. 49ES1807.025, Inservice Inspection Visual Examination Procedure, Rev. 51S1630.410, DGA-T-CTHA DGA Jacket Cooling Water Outlet Temperature Switch Calibration,Rev.91S1630.412, DGA-T-OTHA DGA Lube OilTemperature Switch Calibration, Rev. 8151632.412, DGB-T-OTHA DGB Lube OilTemperature Switch Calibration, Rev. 7MMDI-10, Economy Engineering Hi-Jacker Model HJ-15-E PM and lnspection, Rev. 2MS0523.65, Charging Pump Bearing and Mechanical Seal Maintenance, Rev. 1MX0539.66, B-EDG Mechanical Governor Venting and Setup after Replacement, Rev. 1OS1000.01, Heatup from Cold Shutdown to Hot Standby, Rev. 30OS1000.02, Plant Startup from Hot Standby to Minimum Load, Rev.22OS1000.03, Plant Shutdown from Minimum Load to Hot Standby, Rev. 19OS1000.1 1, Post Trip to Hot Standby, Rev. 8OS1001 .15, PORV Block Valve Operation, Rev. 1OS1016.05, Service Water Cooling Tower Operation, Rev. 20OS1023.75, Operation of Feedwater lsolation Valves'Temporary Heating, Rev. 3OS1026.01, Operation of DG 1,A, Rev. 19OS1026.03, Operating DG 1A Jacket Water Cooling Water System, Rev. 9OS1026.11, Operating DG 1B Jacket Water Cooling Water System, Rev. 9OX1426.01, DG 1A Monthly Operability Surveillance, Rev. 25OX1426.05, DG 1B Monthly Operability Surveillance, Rev.24OX1426.20, Diesel Generator 1A 18 Month Operability and Engineered Safeguards Pump andValve Response Time Testing Surveillance, Rev. 18OX1426.21, Diesel Generator 1B 18 Month Operability and Engineered Safeguards Pump andValve Response Time Testing Surveillance, Rev. 14Attachment A-5OX1426.22, Emergency Diesel Generator 1A24 Hour Load Test and Hot Restart Surveillance,Rev. 13OX1426.23, Emergency Diesel Generator 18 24 Hour Load Test and Hot Restart Surveillance,Rev. 12OX1426.26, DG 1A Semiannual Operability Surveillance, Rev. 14OX1426.27, DG 1B Semiannual Operability Surveillance, Rev. 15OX1426.28, Simultaneous Start of Both Emergency Diesel Generators 1,A and 1B - Ten YearOperability Surveillance Test, Rev. 2OX1431.03, Main Control Valve Quarterly Test, Rev. 24OX1436.07, Main Feedwater System Valve, Cold Shutdown Operability Tests and 18 MonthPosition Verification. Rev. 6OX1456.81, Operability Testing of IST Valves, Rev. 6RS07-01-01, Main Turbine ControlValve Testing, Rev. 1Work Orders00627627 0803990 083682401198138 0803991 083682501198488 0809987 083682701210514 0821860 083682801207753 0836819 08417890803812 0836821 400687170803869 08368220803870 0836823Miscellaneous08-9085599, Platinum (Pt) lncore Detector Assembly, dated 10/18/104.16 kV Distribution System Detailed System Text, Rev. 451-9156705, EQ Review for Seabrook Pt-lCDAs, dated 1Ol18l1O51-9159575. Use of Seabrook Incore S/N PTICD-0006. dated 10118110Ashcroft Reference Bulletin BM-1, BimetalThermometers, Series El, Grade A (1Yo)ASME Section Xl, Ch. 28.38,1WA-5250: Corrective Action, lnterpretation Xl-1-92-31ASTM D6677-01, Standard Test Method for Evaluating Adhesion by KnifeDRR 94-0109, SW Strainer Cover Bolting and Gasket lmprovements, Rev. 3NAH-07-32, Westinghouse Letter - Turbine Valve Testing Data Review, dated 9124107SSPC-SP 12INACE No. 5, Surface Preparation and Cleaning of Metals by Waterjetting Prior toRecoating, July 2002Surveillance and Modification Acceptance Tests1N1652.410, MS-U-500 Steam Dump Control Calibration, completed 4115108KB,V7317, Quality Conformance Test Procedure and Test Documentation for Valve ControlModule 6N372, performed 10/8/091S0563.11, Testing of Agastat 125 VDC (7000 Series) TDPU Timing Relays, performed 1123108and 314108OX1456.81, Operability Testing of IST Valves: Valve Stroke Time and Remote Position/StatusLight Data: 1-FW-V-30, performed 1121108UT Examination Report: SW-1 81 4-1 -1 56-24, performed I 0127 lQ9Attachment A-6Vendor ManualsFP25277. DG Governor Modification lnstruction Manual, Rev. 2FP52936, Charging/Safety Injection Pumps Operations Manual, Rev. 23FP53455, Primary Component Cooling Pump, Rev. 11Mobil DTE 700 Series, Product Specifications and Properties, dated 6/07Mobil DTE 790 Series, Product Specifications and Properties, dated 11106PDS l-06. Mobil Product Data Sheet - Mobil DTE 797 OitADAMSASMECBSCFRCOLRCRDRSEDGFlVsFSARrMcLARMSIVNCVNEINRCNRRPARSPMTRAIRATSDPSFPSISLIVSWTSUATUFSARLIST OF ACRONYMSAgencywide Documents Access and Management SystemAmerican Society of Mechanical EngineersContainment Building SprayCode of Federal RegulationsCore Operating Limits ReportCondition ReportsDivision of Reactor SafetyEmergency Diesel GeneratorFeed Water lsolation ValvesFinal Safety Analysis Reportlnspection Manual ChapterLicense Amendment RequestMain Steam lsolation ValveNon-Cited ViolationNuclear Energy InstituteNuclear Regulatory CommissionOffice of Nuclear Reactor RegulationPublicly Available RecordsPost Modification TestRequest for Additional InformationReserve Auxiliary TransformerSignificance Determination ProcessSpent Fuel PoolSafety lnjectionSeverity Level lVService WaterTechnical SpecificationsUnit Auxiliary TransformerUpdated Final Safety Analysis ReportAttachment