IR 05000333/2011005

IR 0500033312011005; 10/01/2011 - 12/31/2011; James A. FitzPatrick Nuclear Power Plant (FitzPatrick); Emergent Work Control, ALAM Planning and Controls, and Follow-up of Events

ML12038A231
Person / Time
Site:	FitzPatrick
Issue date:	02/07/2012
From:	Mel Gray Reactor Projects Branch 2
To:	Michael Colomb Entergy Nuclear Northeast
Gray M
References
Download: ML12038A231 (39)

Text

with a copies to the Regional Administrator, Region l; theDirecior, Office of Enforcement; United States Nuclear Regulatory Commission, Washington,DC 20555-0001; and the NRC Senior Resident Inspector at FitzPatrick. ln addition, if youdisagree with the cross-cutting aspect assigned to any finding in this report, you should providea reJponse within 30 days of the date of this inspection report, with the basis for yourdisagreement, to the RegionalAdministrator, Region l, and the NRC Senior Resident Inspectorat FitzPatrick.

M. ColombIn accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letter andits enclosure, and your response (if any)will be available electronically for public inspection inthe NRC Public Document Room or from the Publicly Available Records (PARS) component ofthe NRC's document system (ADAMS). ADAMS is accessible from the NRC Website athttp://www.nrc.gov/reading-rm/adams.htm! (the Public Electronic Reading Room).

Sincerely,ru-(Mel Gray, QhiefDivision of Reactor ProjectsDocket No.:License No.:

Enclosure:

cc Mencl:50-333DPR-59lnspection Report 05000333/201 1 005

w/Attachment:

Supplementary InformationDistribution via ListServ 2M. Colombln accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of this letter andits enclosure, and your response (if any) will be available electronically for public inspection inthe NRC Public Document Room or from the Publicly Available Records (PARS) component ofthe NRC's document system (ADAMS). ADAMS is accessible from the NRC Website athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RAMel Gray, ChiefReactor Projects Branch 2Division of Reactor ProjectsDocket No.: 50-333License No.: DPR-59

Enclosure:

lnspectionReport05000333/2011005

w/Attachment:

Supplementary I nformationDistribution (via e-mail):W. Dean, RAD. Lew, DRAJ. Tappert, DRPJ. Clifford, DRPC. Miller, DRSP. Wilson. DRSM. Gray, DRPB. Bickett, DRPS. McCarver, DRPM. Jennerich, DRPE. Knutson, DRP, SRIB. Sienel, DRP, RlK. Kolek, Resident OAL. Chang, OEDORidsNrrPMFitzPatrick ResourceRidsNrrDorlLpll -1 ResourceROPreports ResourceSUNSI Review Complete: p! (Reviewer's Initials)MLl2038A231DOCUMENT NAME: G:\DRP\BRANCH2\a - Fitzpatrick\Reports\Fitr lR 2011 005\FiEP lR 2011 005 Final.docxAfter declaring this document "An Official Agency Record" it will be released to the Public.To receive a copy of this document, indicate in the box: "C" = Copy without attachmenUenclosure "E" = Copywith attachmenUenclosure "N" = No copy Docket No.:License No.:Report No.:Licensee:Facility:Location:Dates:Inspectors:Approved by:1U.S. NUCLEAR REGULATORY COMMISSIONREGION I50-333DPR-5905000333/201 1 005Entergy Nuclear Northeast (Entergy)James A. FitzPatrick Nuclear Power PlantScriba, New YorkOctober 1 through December 31,2011E. Knutson, Senior Resident InspectorB. Sienel, Resident InspectorB. Bickett, Senior Project EngineerS. McCarver, Project EngineerR. Rolph, Health PhysicistMel Gray, ChiefReactor Projects Branch 2Division of Reactor ProjectsEnclosure 2

SUMMARY OF FINDINGS

............ .........3REPORT DETATLS ..,............61. REACTOR SAFETY .......................61R01 Adverse Weather Protection. ...........,.....61R04 Equipment Alignment ...........61R05 Fire Protection........... ...........71R07 Heat Sink Performance .............. ...........81R11 Licensed Operator Requalification Program .........,.... ..............91R12 Maintenance Effectiveness.......... ..........91R13 Maintenance Risk Assessments and Emergent Work Control ................101R15 Operability Determinations and Functionality Assessments.......... ..........121R18 Plant Modifications .............131R19 Post-Maintenance Testing..... ..............131R22 Surveillance Testing ......,....141EPO Drill Evaluation........... ........152. RADlATlON SAFETY ...................162RS2 OccupationalALAM Planning and Controls........,..... .."'.....'162RS3 In-Plant Airborne Radioactivity Control and Mitigation............ ..........".".194. OTHER ACTtVtrES..........,.... ......204OA1 Performance Indicator Verification "'.'.204OA2 Problem ldentification and Resolution '.....'.......'...214OA3 Follow-up of Events and Notices of Enforcement Discretion.. ........"""".244OAO Meetings, Including Exit.......... .'...."'...284OAT Licensee'ldentified Violations. "..'...'.".28ATTACHMENT: SUPPLEMENTARY INFORMATION ................. '."'."..28SUPPLEMENTARY INFORMATION.......... ........A-1KEy potNTS OF CONTACT ...,.,....... A-1LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED .... A-1LIST OF DOCUMENTS REVIEWED ........... .,.,.,,4.2Ltsr oF ACRoNYMS............... ......... A-7Enclosure 3SUMMARY OF FINDINGSf R 0500033312011005; 10i01/2011 - 1213112011; James A. FitzPatrick Nuclear Power Plant(FitzPatrick); Emergent Work Control, ALAM Planning and Controls, and Follow-up of Events.The report covered a three-month period of inspection by resident inspectors and announcedinspections by region-based inspectors. Four Green findings, three of which were non-citedviolations (NCVs), were identified. The significance of most findings is indicated by their color(Green, White, Yellow, Red) using Inspection Manual Chapter (lMC) 0609, "SignificanceDetermination Process" (SDP). The cross-cutting aspects for the findings were determinedusing IMC 0310, "Components Within the Cross-Cutting Areas." Findings for which the SDPdoes not apply may be "Green" or be assigned a severity level after Nuclear RegulatoryCommission (NRC) management review. The NRC's program for overseeing the safe operationof commercial nuclear power reactors is described in NUREG-1649, "Reactor OversightProcess," Revision 4, dated December 2006.

Cornerstone: Mitigating Systemso

Green.

The inspectors identified a non-cited violation (NCV) of Technical Specification (TS)3.3.1.1, "Reactor Protection System (RPS) Instrumentation," because FitzPatrick operatorsdid not take required action within the allowed completion time in response to an RPS relayfailure. Specifically, following failure of RPS channel 'B' shutdown scram reset interlocklogic relay 5A-K178, which caused the reactor mode switch to shutdown manual scram tobe disabled, action was not taken by operators to insert a half-scram on RPS channel 'B'within one hour as required by TS 3.3.1.1 Condition C. After further evaluation of the issue,operators inserted a half scram on RPS channel 'B'. The issue w"as entered into thecorrective action program (CAP) as condition report (CR)-JAF-2O11-06625.The finding was more than minor because it affected the equipment performance attribute ofthe Mitigating Systems cornerstone objective to ensure the availability of systems thatrespond to initiating events to prevent undesirable consequences. Specifically, the delay inimplementing the TS required actions resulted in additional accrual of more than two hoursof reactor operation with the reactor mode switch to shutdown manual scram bypassed.The inspectors evaluated the finding using the Phase 1, "lnitial Screening andCharacterization of Findings," worksheet in Attachment 4 to IMC 0609, "SignificanceDetermination Process." The inspectors determined this finding was not a designqualification deficiency resulting in a loss of functionality or operability, did not represent anactual loss of safety function of a system or train of equipment, and was not potentially risksignificant due to external initiating events. Therefore, the inspectors determined the findingto be of very low safety significance (Green). This finding had a cross-cutting aspect in thearea of Human Performance, decision making, because operators did not use conservativeassumptions in decision making and promptly apply readily available information containedin the alarm response procedure and TS Bases to determine TS applicability for the alarmcondition tH.1(b) per lMC03101. (Section 1R13).

Green.

The inspectors identified a self-revealing NCV of 10 CFR 50, Appendix B, CriterionXVl, "Corrective Action," because Entergy personnel did not promptly correct the intermittentfailure of reactor core isolation cooling (RCIC) steam admission valve 13MOV-131 to fullyopen on demand. Specifically, Entergy staff's troubleshooting performed in response to theOctober 29,2010, partial valve opening was not adequate in scope to identify the cause ofEnclosure 4the intermittent failure. As corrective action, a more extensive troubleshooting effort wasundertaken by Entergy staff following a second failure of the valve to fully open onJanuary 7,2011, which was successful at identifying and correcting the problem. The issuewas entered into the CAP as CR-JAF-2011-00123.The finding was more than minor because it affected the equipment performance attribute ofthe Mitigating Systems cornerstone objective to ensure reliability of systems that respond toinitiating events to prevent undesirable consequences. Specifically, the loose electricalconnections in the 13MOV-131 motor control circuit affected the reliability of the RCICsystem. Since the RCIC pump achieved rated discharge flow and pressure on bothoccasions that 13MOV-131 failed to fully open, the inspectors concluded that RCICremained capable of performing its design function during the period that this conditionexisted. The inspectors evaluated the finding using the Phase 1, "lnitial Screening andCharacterization of Findings," worksheet in Attachment 4 to IMC 0609, "SignificanceDetermination Process." The inspectors determined this finding was not a designqualification deficiency resulting in a loss of functionality or operability, did not represent anactual loss of safety function of a system or train of equipment, and was not potentially risksignificant due to external initiating events. Therefore, the inspectors determined the findingto be of very low safety significance (Green). The finding had a cross-cutting aspect in thearea of Human Performance, work control, because Entergy personnel did not appropriatelyplan the scope of 13MOV-131 troubleshooting activity by incorporating consideration of thehigh risk significance of the RCIC system H.3(a) per lMC0310]. (Section 4OA3)

Cornerstone: Radiation Safetyr

Green.

The inspectors identified a self-revealing finding that involved inadequate workplanning relative to the 'A' recirculation pump replacement work during refueling outageR19 that resulted in additional unplanned collective exposure (39.168 person-rem comparedto a work activity estimate of 15.831 person-rem). The actual job site conditions were notadequately evaluated by Entergy staff for interferences and the support work was notcoordinated to prevent additional unnecessary exposure and did not meet the RadiationWork Permit (RWP) No. 10-0518 planned dose execution for the work activity. Thisinadequate evaluation lead to as-found interferences that required removal andreinstallation, and insufficient outage schedule coordination that resulted in several scaffoldinterferences with other outage tasks that caused avoidable scaffold rework and inunintended exposure that could have been avoided by Entergy personnel.The finding was more than minor because it was associated with the Radiation Safety -Occupational Radiation Safety cornerstone attribute of program and process, and affectedthe cornerstone objective of protecting worker health and safety from exposure to radiation.Specifically, inadequate work planning resulted in unplanned, unintended collectiveexposure that was greater than 50 percent above the intended collective exposure andgreater than five person-rem due to conditions that were reasonably within Entergy's abilityto foresee and correct. The inspectors evaluated the finding using IMC 0609, Appendix C,"Occupational Radiation Safety Significance Determination Process," and determined thatthe finding was of very low safety significance (Green) because the finding was due toAs Low As Reasonably Achievable (ALARA) work control planning and the three year rollingEnclosure 5average collective exposure at FitzPatrick was less than 240 person-rem (146.593 person-rem for 2008-2010). The finding had a cross-cutting aspect in the area of HumanPerformance, work control, because Entergy's planned work activities did not adequatelyincorporate work site interferences or outage work coordination in the work control planningprocess H.3(b) per lMC0310]. (Section 2RS2).

Green.

The inspectors identified a self-revealing NCV of TS 5.4, "Procedures," whichrequires that written procedures be implemented covering the activities in the applicableprocedures recommended by Regulatory Guide 1.33, including procedures for RWPs andALARA reviews. Specifically, as of December 12, 2011, post job reviews for most of the2010 R-19 RWPs (52 of 55) had not been completed as required by procedure EN-RP-105,"RadiologicalWork Permits," Revision 10. This procedure requires post job reviews to becompleted within 90 days from the end of the outage. The performance deficiency couldlead to repeating errors and not planning the upcoming R-20 with needed improvements.Since planning for the R-20 outage had already begun, the inspectors concluded thatlessons learned in the R-19 outage RWPs may not be incorporated into the R-20 RWPs andadditional, avoidable exposure could be received. Entergy staff subsequently developed atracking schedule to complete the reviews and entered the issue into the CAP as CR-JAF-2011-04152.The finding was more than minor because it was associated with the Radiation Safety -Occupational Radiation Safety cornerstone attribute of program and process, and affectedthe cornerstone objective of protecting worker health and safety from exposure to radiation.Specifically, Entergy staff did not complete RWP close out documentation to identify lessonslearned and actions to reduce worker exposure in subsequent refueling outages. Theinspectors evaluated the finding using IMC 0609, Appendix C, "Occupational RadiationSafety Significance Determination Process," and determined that the finding was of very lowsafety significance (Green) because it did not involve: (1) ALARA planning and controls,(2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability toassess dose. The finding had a cross-cutting aspect in the area of Human Performance,work practices, because Entergy personnel did not effectively communicate expectationsregarding procedural compliance tH.4(b) per lMC0310]. (Section 2RS2)

Other Findings

A violation of very low safety significance that was identified by Entergy personnel wasreviewed by the inspectors. Corrective actions taken or planned by Entergy personnel havebeen entered into FitzPatrick's corrective action program. This violation and correctiveaction tracking number are listed in Section 4OA7 of this report.Enclosure 6

REPORT DETAILS

Summarv of Plant StatusThe James A. FitzPatrick Nuclear Power Plant (FitzPatrick) began the inspection period at100 percent power. On November 14, 2011 , operators reduced reactor power to 50 percent torepair several steam leaks on balance of plant equipment and to perform control rod bladeinterference monitoring. Operators returned the plant to 100 percent power on the same day.On December 2, 2011, operators reduced power to 50 percent to plug leaking main condensertubes. Operators returned the plant to 100 percent power on December 4, 2011. OnDecember 27,2011, operators reduced power to 50 percent to plug leaking main condensertubes, to perform a control rod sequence exchange, and to perform control rod bladeinterference testing. Operators returned the plant to 100 percent power on December 29, 2011,and remained at or near 100 percent power for the remainder of the inspection period.1.

REACTOR SAFETY

Gornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01 Adverse Weather Protection (71111.01 - 1 sample)a. Inspection ScopeThe inspectors performed a review of the station's readiness for the onset of seasonallow temperatures. The review focused on the emergency diesel generator (EDG), 125volt direct current (VDC) battery, and standby gas treatment systems. The inspectorsreviewed the Updated Final Safety Analysis Report (UFSAR), TSs, control room logs,and the CAP to determine what temperatures or other seasonal weather could challengethese systems, and to ensure Entergy personnel had adequately prepared for thesechallenges. The inspectors reviewed station procedures, including the seasonalweatherpreparation procedure and applicable operating procedures. The inspectors performedwalkdowns of the selected systems to ensure station personnel had identified issues thatcould challenge the operability of the systems during cold weather conditions.Documents reviewed for each section of this inspection report are listed in theAttachment.These activities constituted one seasonal weather conditions inspection sample.b. FindinqsNo findings were identified.

1R04 Equipment Alionment

.1 Partial Svstem Walkdown (71111.04Q - 3 samples)Enclosure 7a. Inspection ScopeThe inspectors performed partial walkdowns of the following systems:o 'A' and 'C' EDGs during 'B' EDG maintenance on October 26,2Q11r 'B' and 'D' EDGs during 'C' EDG maintenance on November 15,2011. RCIC system following system maintenance on December 15,2011The inspectors selected these systems based on their risk-significance relative to thereactor safety cornerstones at the time they were inspected. The inspectors reviewedapplicable operating procedures, system diagrams, the UFSAR, TSs, work orders(WOs), and CRs, in order to identify conditions that could have impacted systemperformance of their intended safety functions. The inspectors performed fieldwalkdowns of accessible portions of the systems to verify system components andsupport equipment were aligned correctly and were operable. The inspectors examinedthe material condition of the components and observed operating parameters ofequipment to verify that there were no deficiencies. The inspectors also reviewedwhether Entergy staff had identified equipment issues and entered them into the CAP forresolution with the appropriate significance characterization, as required by 10 CFR 50,Appendix B, Criterion XVl, "Corrective Action."These activities constituted three partial system walkdown inspection samples.b. FindinqsNo findings were identified.

1R05 Fire Protection

.1 Resident Inspector Quarterlv Walkdowns (71111.05Q - 5 samples)a. Inspection ScopeThe inspectors conducted tours of the areas listed below to assess the materialcondition and operational status of fire protection features. The inspectors verified thatstation personnel controlled combustible materials and ignition sources in accordancewith administrative procedures. The inspectors verified that fire protection andsuppression equipment were available for use as specified in the area pre-fire plan, andpassive fire barriers were maintained in good material condition. The inspectors alsoverified that station personnel implemented compensatory measures for out of service,degraded, or inoperable fire protection equipment, as applicable. The inspectorsevaluated the fire protection program for conformance with the requirements of licensecondition 2.C (3), "Fire Protection.". 'B' and 'D' EDG rooms and switchgear room, fire arealzone VI/EG-3, EG-4, EG-6, onNovember 15,2011o 'A' and 'B' station battery room complex, fire arealzone lll/BR-1, BR-2, lV/BR-3,BR-4, XVIiBR-S, on November 15,2011Enclosure

1R07 Heat Sink Performance (71111.07 - 1 sample)a. lnspection ScopeThe inspectors reviewed the results of the 'B' EDG jacket water cooler inspection thatwas performed by Entergy staff on October 26, 2011. This heat exchanger is cooled bythe emergency service water system. The inspectors also discussed the results of thisand past jacket water cooler inspections as well as the frequency of inspections with theservice water system and heat exchanger program engineers to verify the inspectionfrequency was appropriate.These activities constituted one heat sink performance inspection sample.Enclosure

9b. FindinqsNo findings were identified.

1R11 Licensed Operator Reoualification Prooram (71111.11Q - 1 sample)a. Inspection ScopeThe inspectors observed licensed operator simulator training on November 16, 2011,which included simulated loss of a Division 2 motor control center, an isolable steamleak from the high pressure coolant injection (HPCI) system, loss of a condensate pump,a manual scram with failure of all control rods to insert, loss of all feed and condensate,and an emergency reactor pressure vessel depressurization. The inspectors evaluatedoperator performance during the simulated event and verified completion of risksignificant operator actions, including the use of abnormal and emergency operatingprocedures. The inspectors assessed the clarity and effectiveness of communications,implementation of actions in response to alarms and degrading plant conditions, and theoversight and direction provided by the control room supervisor. The inspectors verifiedthe accuracy and timeliness of the emergency classifications made by the shift manager.Additionally, the inspectors assessed the ability of the crew and training staff to identifyand document crew performance problems.These activities constituted one quarterly operator simulator training inspection sample.b. FindinqsNo findings were identified,1R12 Maintenance Effectiveness (71111JzQ - 2 samples)a. Inspection ScopeThe inspectors reviewed the samples listed below to assess the effectiveness ofmaintenance activities on structures, systems, and components (SSCs) performanceand reliability. The inspectors reviewed system health reports, corrective action programdocuments, maintenance work orders, and maintenance rule basis documents to ensurethat Entergy personnelwere identifying and properly evaluating performance problemswithin the scope of the maintenance rule. For each sample selected, the inspectorsverified that the SSC was properly scoped into the maintenance rule in accordance with10 CFR 50.65 and verified that the (a)(2) performance criteria established by staff werereasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed theadequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally,the inspectors ensured that Entergy staff were identifying and addressing commoncause failures that occurred within and across maintenance rule system boundarieswhen applicable.. Containment air dilutiono EDGThese activities constituted two quarterly maintenance effectiveness inspection samples.Enclosure

I10b. FindinqsNo findings were identified.inf g Maintenance RiskAssessments and EmerqentWork Control (71111.13-4samples)a. Inspection ScopeThe inspectors reviewed maintenance activities to verify that the appropriate riskassessments were performed prior to removing equipment for work. The inspectorsreviewed whether risk assessments were performed as required by 10 CFR 50.65(a) (4),and were accurate and complete. When emergent work was performed, the inspectorsreviewed whether plant risk was promptly reassessed and managed. The inspectorsalso walked down selected areas of the plant which became more risk significantbecause of the maintenance activities to ensure they were appropriately controlled tomaintain the expected risk condition. The reviews focused on the following activities:. Planned maintenance of the 'B' EDG during the week of October 24,2011r Planned outage of 1 15 kilovolt (kV) offsite power line 4 during the week ofOctober 31, 2011o Planned maintenance of 'C' EDG, steam leaks in the balance of plant, and controlblade interference monitoring during the week of November 14,2011. Planned maintenance of 'B'and 'D' EDG, HPCI systeffi, 'B' residual heat removal(RHR) system, and emergent maintenance to replace a 'B' reactor protection system(RPS) relay and repair a leak from the turbine building closed loop cooling systemduring the week of December 19,2011These activities constituted four maintenance risk assessments and emergent workcontrol inspection samples.b. FindinssIntroduction: The inspectors identified an NCV of very low safety significance (Green) ofTS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," because FitzPatrickoperators did not take required action within the allowed TS completion time in responseto an RPS relay failure. Specifically, following failure of RPS channel 'B' shutdownscram reset interlock logic relay 5A-K178, which caused the reactor mode switch toshutdown manual scram to be disabled, action was not taken by operators to insert ahalf-scram on RPS channel 'B'within one hour as required by TS 3.3.1,1 Condition C.Description: At 8:05 p.m. on December 19,2011, control room annunciator 09-5-1-33,"Mode Switch in Shutdown Trip Bypassed," alarmed. Operators began investigation ofthe cause and reviewed the TS to determine what actions were required. Operatorsdetermined that RPS channel 'B'should be considered inoperable, and therefore, thatTS 3.3.1,1, "Reactor Protection System (RPS) lnstrumentation," Condition A, "One ormore required channels inoperable," applied. The required action for this condition wasto place the associated trip system in trip, with an allowed completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.Enclosure 11Operators did not immediately trip RPS channel 'B' because, in part, the resultant halfscram would place the plant in a less reliable condition, in that a single spurious tripsignal to the remaining RPS channel (channel 'A') would cause an unnecessary reactorscram.The reactor mode switch is a four position switch that controls the operating modes ofthe RPS. One of the functions of the reactor mode switch in the "shutdown" position isto initiate a manual reactor scram. Since placing the reactor mode switch in shutdown isan immediate operator action following a scram, this function serves to enforce thesignal that produced the scram. This scram signal is automatically bypassed by designtwo seconds after it is applied, to allow the scram to be manually reset (i.e., allow thetwo RPS channels to be reenergized). This action is desirable because it restores thenormal valve lineup in the control rod drive hydraulic system, which stops water frombeing ported through the control rod drive mechanisms into the reactor vessel.During Entergy staff's review of the condition, operators determined that RPS channel'B'shutdown scram reset interlock logic relay 5A-K178 was deenergized and appearedto have burned out. At approximately 11:00 p.m., operators realized that this conditioncaused the reactor mode switch to shutdown manual scram to be inoperable; that wasbecause the bypass had been applied to RPS channel 'B'when 5A-K178 failed,therefore the reactor mode switch to shutdown manual scram would only result in a halfscram on RPS channel 'A'. Since TS Table 3.3.1.1-1 includes the reactor mode switchto shutdown manual scram as a required trip function, TS 3.3.1.1 Condition C, "One ormore functions with RPS trip capability not maintained," also applied, with an allowedcompletion time of one hour. At 1 1 :17 p.m. on December 19, 2011 , three hours and12 minutes after the alarm had occurred, operators inserted a half scram on RPSchannel 'B'.During inspection of this issue, the inspectors reviewed the TS basis for the reactormode switch to shutdown manual scram, which states, "The reactor mode switch willscram the reactor if it is placed in the shutdown position . . . Two channels of ReactorMode Switch - Shutdown Position Function, with one channel in each trip system, areavailable and required to be Operable." The inspectors also reviewed the applicablealarm response procedure (ARP) 09-5-1-33, "Mode SW [switch] in Shutdown TripBypassed," Revision 2. The inspectors noted that the "Automatic Actions" in theprocedure stated, "Reactor mode switch to shutdown manual scram is bypassed." Theinspectors considered that this information should have been immediately available tothe operators, because review of the applicable ARP is the expected initial response toany alarm condition. Since the TS bases indicated that the reactor mode switch toshutdown manual scram is a reactor trip function, and the ARP specifically stated theplant condition associated with the alarm, the inspectors concluded that there had beenadequate information readily available to the operators to have determined that TS3.3.1.1 Condition C was applicable, and to have inserted a half scram on RPS channel'B' within one hour of the alarm having occurred.The inspectors noted that, if the TS 3.3.1.1 Condition C required action and associatedcompletion time were not met, TS 3.3.1.1 Condition G required that the plant be inMode 3 (hot shutdown) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Although this completion time had beenEnclosure 12satisfied, the inspectors also noted that there had been no reason that the requiredaction of Condition C (to insert a half scram) could not be performed. The inspectorsdiscussed their conclusions with FitzPatrick management and the issue was entered intothe CAP as CR-JAF-2011-06625 for further evaluation.Analysis: The inspectors determined that control room operators not promptly enteringTS 3.3.1.1 Condition C in response to the "mode switch in shutdown trip bypassed"alarm was a performance deficiency. The finding was more than minor because itaffected equipment performance attribute of the Mitigating Systems cornerstoneobjective to ensure the availability, reliability, and capability of systems that respond toinitiating events to prevent undesirable consequences. Specifically, it resulted inadditional accrual of more than two hours of reactor operation with the reactor modeswitch to shutdown manual scram bypassed. The inspectors evaluated the finding usingthe Phase 1, "lnitial Screening and Characterization of Findings," worksheet inAttachment 4 to IMC 0609, "Significance Determination Process." The inspectorsdetermined this finding was not a design qualification deficiency resulting in a loss offunctionality or operability, did not represent an actual loss of safety function of a systemor train of equipment, and was not potentially risk significant due to a seismic, fire, orsevere weather initiating event. Therefore, the inspectors determined the finding to be ofvery low safety significance (Green).This finding had a cross-cutting aspect in the area of Human Performance, decisionmaking, because operators did not use conservative assumptions in decision makingand promptly apply readily available information contained in the alarm responseprocedure and TS Bases to determine TS applicability for the alarm condition tH.1(b)1.Enforcement: TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," ConditionC, "One or more Functions with RPS trip capability not maintained," while the plant is inMode 1 or 2, requires that the RPS trip capability be restored within one hour, TSTable 3.3.1 ,1-1 , "Reactor Protection System Instrumentation," Function 10 identifies the,"Reactor mode switch - shutdown position," as a required function. Contrary to theabove, following a loss of the reactor mode switch to shutdown manual scram functiondue to failure of RPS channel 'B' shutdown scram reset interlock logic relay 5A-K178 at8:05 p,m. on Decemberlg, 2011, whilethe plantwas in Mode 1, operatorsdid notrestore the affected RPS trip capability by inserting a half scram on RPS channel 'B' until11:17 p.m., December 19, 2011, a period in excess of three hours, despite theavailability of information to operators to support completion within one hour. Becausethis issue is of very low safety significance (Green) and Entergy personnel entered thisissue into their CAP as CR-JAF-2011-06625, this finding is being treated as an NCVconsistent with the NRC Enforcement Policy. (NCV 05000333/2011005-01, ModeSwitch in Shutdown Scram Function Inoperable in Excess of the TS AllowedOutage Time due to Personnel Error)1 R15 Operabilitv Determinations and Functionalitv Assessments (71111.15 - 3 samples)a. Inspection ScopeThe inspectors reviewed operability determinations for the following degraded or non-conforming conditions:Enclosure 13. CR-JAF-2U1-A5372 regarding cover gasket leakage on the 'A' RHR service waterstrainer, 10S-5A1, on October 20, 2011. CR-JAF-2011-06067 concerning the operability of 'D' EDG while the fuel oil transferpump selector switch was mispositioned such that the lead pump would not startwhen required, on November 21,2011e CR-JAF-2011-06536 regarding 24VDC battery charger 7118C-4low voltage alarmon December 15,2011The inspectors selected these issues based on the risk significance of the associatedcomponents and systems. The inspectors evaluated the technical adequacy of theoperability determinations to assess whether TS operability was properly justified andthe subject component or system remained available such that no unrecognizedincrease in risk occurred. The inspectors compared the operability and design criteria inthe appropriate sections of the TS and UFSAR to Entergy personnel's evaluations todetermine whether the components or systems were operable. Where compensatorymeasures were required to maintain operability, the inspectors determined whether themeasures in place would function as intended and were properly controlled by Entergypersonnel. The inspectors determined, where appropriate, compliance with boundinglimitations associated with the evaluations.These activities constituted three operability evaluation inspection samples.b. FindinqsNo findings were identified.

1R18 Plant Modifications (71111.18 - 1 sample)a. lnspection ScopeThe inspectors reviewed the temporary modification listed below to determine whetherthe modification affected the safety functions of systems that are important to safety.The inspectors reviewed 10 CFR 50.59 documentation and post-modification testingresults and verified that the temporary modification did not degrade the design basis,licensing basis, and performance capability of the affected system.. Engineering Change (EC) 30962, Temporary Alarm Setpoint Change for 20TlS-5348Reactor Building Equipment Sump Temp.These activities constituted one temporary modification inspection sample.b. FindinqsNo findings were identified.1R19 Post-Maintenance Testinq (71111.19 - 5 samples)Enclosure

14a. lnspection ScopeThe inspectors reviewed post-maintenance tests (PMTs) for the maintenance activitieslisted below to verify that procedures and test activities ensured system operability andfunctional capability. The inspectors reviewed the test procedures to verify that theyadequately tested the safety functions that may have been affected by the maintenanceactivity, that the acceptance criteria in the procedure was consistent with the informationin the applicable licensing basis and/or design basis documents (DBDs), and that theprocedure had been properly reviewed and approved. The inspectors also witnessedthe test or reviewed test data to verify that the test results adequately demonstratedrestoration of the affected safety functions.o WO 00283141 to replace the 4 kV emergency bus 10600 degraded voltage timedelay relay, on October 28,2011. WO 00279627 to replace the -15.00 VDC power supply for stack radiation monitor17RM-53A, on November 9, 2011o WO 52040138-06 to replace EDG vent system C exhaust system damper operatorsOP1, OP2, OP3, and OP4, on November 18,2011. WO 297765-04 to troubleshoot failure of 'D' EDG fuel oil transfer pump 93P1-D1 tooperate prior to receipt of the day tank low level alarm, on November 22,2011. WO 00291 1 19 to open, inspect, and replace valve internals for RCIC full flow test tocondensate storage tanks check valve, on December 14,2011These activities constituted five PMT inspection samples.b. FindinqsNo findings were identified. '1R22 Surveillance Testino (71111.22 - 5 samples)a. lnspection ScopeThe inspectors witnessed performance of surveillance tests (STs) and/or reviewed testdata of selected risk-significant SSCs to assess whether the SSCs satisfied TSs,UFSAR, technical requirements manual, and station procedure requirements. Theinspectors verified that test acceptance criteria were clear, demonstrated operationalreadiness and were consistent with DBDs, test instrumentation had current calibrationsand the range and accuracy for the application, tests were performed as written, andapplicable prerequisites were satisfied. Upon ST completion, the inspectors verified thatequipment was returned to the status specified to perform its safety function. Theinspectors reviewed the following STs:. ST-3PA, "Core Spray Loop A Quarterly Operability Test (lST [in-service test]),"Revision 20, on November 10,2011. SP-01.02, "Reactor Water Sampling and Analysis," Revision 23, on November 17,2011. ST-6H8, "Standby Liquid Control B Side Quarterly Operability Test (lST),"Revision 6, on November 23,2011Enclosure 15. ISP-94A-MG, "Reactor Protection System Electrical Protection Assembly'A" MG[motor-generator] FunctionaliCalibration," Revision 4, on December 9, 2011. ISP-23A, "Emergency Service Water Lockout Matrix Instrument ChannelCalibration," Revision 2, on December 13,2011These activities represented five surveillance testing inspection samples.b. FindinqsNo findings were identified.Cornerstone: Emergency Preparednessl EPO Drill Evaluation (711 14.06 - 2 samples).1 Emerqencv Preparedness Drill Observationa. Inspection ScopeThe inspectors evaluated the conduct of a routine FitzPatrick emergency drill onNovember 30, 2011, to identify weaknesses and deficiencies in the classiflcation,notification, and protective action recommendation development activities. Theinspectors observed emergency response operations in the simulator and, technicalsupport center (TSC), and emergency operations facility to determine whether the eventclassifications, notifications, and protective action recommendations were performed inaccordance with procedures, The inspectors also attended the TSC drill critique tocompare inspector observations with those identified by Entergy staff in order toevaluate the staffs critique and to verify whether Entergy staff were properly identifyingweaknesses and entering them into the CAP,These activities represented one drill evaluation inspection sample.b. FindinosNo findings were identified..2 Traininq Observationa. Inspection ScopeThe inspectors observed a simulator training evolution for FitzPatrick licensed operatorson June 20,2011 (omitted in error from lnspection Report 0500033312011003), whichrequired emergency plan implementation by an operations crew. Entergy staff plannedfor this evolution to be evaluated and included in performance indicator (Pl) dataregarding drill and exercise performance. The inspectors observed event classificationand notification activities performed by the crew. The inspectors also attended the post-evolution critique for the scenario. The focus of the inspectors' activities was to note anyEnclosure b.16weaknesses and deficiencies in the crew's performance and ensure that Entergyevaluators noted the same issues and entered them into the CAP.These activities represented one simulator training evaluation inspection sample.FindinqsNo findings were identified.RADIATION SAFEWCornerstones: Occupational Radiation Safety and Public Radiation SafetyOccupationalALARA Planninq and Controls (71124.02 - 1 sample)Inspection ScopeDuring the period from December 12,2011, through December 15,2011, the inspectorsconducted the following activities to verify that Entergy staff was properly implementingoperational, engineering, and administrative controls to maintain personnel exposureALARA. lmplementation of these controls was reviewed against the criteria contained in10 CFR Part 20, applicable industry standards, and station procedures.Radioloqical Work Planninoo The inspectors obtained a list of the work activities ranked by estimated exposure forthe most recent refueling outage, R-19 (2010).. The inspectors reviewed the ALARA work activity evaluations, exposure estimates,and exposure control requirements.. The inspectors verified Entergy staff identified appropriate dose mitigation, definedreasonable dose goals, included decreased worker efficiency from use of respiratorsand heat stress, and included remote technologies.. The inspectors compared the actualexposure received with the dose estimates andthe actual hours with the estimated hours.Verification of Dose Estimates and Exposure Trackinq Svstemso The inspectors reviewed the assumptions and basis described in the R-19 RWP andALARA packages for'A' reactor recirculating pump replacement, safety relief valvework, reactor disassembly/reassembly, in-service inspection activities, and refuelingactivities. The inspectors reviewed the "ALAM'and the "ALARA and RWPPreparation" procedures to determine Entergy staff's methodology for estimatingexposures for specific work activities.r The inspectors verified, for the above activities, that Entergy staff had establishedmeasures to track, trend, and adjust occupational dose estimates for ongoing workactivities. The inspectors verified trigger points were used to prompt additionalreviews.o The inspectors reviewed Entergy staff's method for adjusting exposure estimateswhen unexpected changes in scope, dose rates, or emergent work wereencountered.2.2RS2a.Enclosure 17Problem ldentification and Resolutiono The inspectors verified that problems associated with ALARA planning and controlswere identified in the CAP and were properly addressed.b. Findinqs.1 Inadequate Work Planninq for'A' Reactor Recirculation Pump Replacementlntroduction: The inspectors identified a self-revealing finding of very low safetysignificance (Green) because Entergy personnel did not adequately plan and coordinateR-19 work activities to prevent unnecessary exposure consistent with the original doseestimate as described in RWP No. 10-0518. Specifically, inadequate work planning andcoordination issues relative to the reactor recirculation pump replacement resulted in anunplanned collective exposure of 39.168 person-rem compared to an original workestimated dose of 15.831 person'rem.Description: FitzPatrick RWP No. 10-0518 provided the applicable plan for doseexecution related to the 'A' reactor recirculation pump replacement work during R-19.The activity was planned by Entergy personnel prior to the refueling outage using thenormal outage planning and scheduling process. The inspectors determined the actualversus planned job site conditions were not adequately evaluated by Entergy personnelfor interferences and support work involving scaffolding. Specifically, the inspectorsdetermined there was a lack of in-field walkdowns by Entergy staff prior to the activitythat resulted in unidentified interferences. As a result, Entergy staff received additionalunnecessary exposure.The inspectors determined that in-field high radiation work resulted in additionalcollective exposure that could have been avoided if station personnel had performedsufficient work activity planning and radiation protection had stopped high radiation workuntil project management provided updated work status and coordination. Theinspectors determined the actual work activity exposure of 39,168 person-rem was147 percent greater than the original estimate of 15.831 person-rem for the 'A' reactorrecirculation pump replacement. Entergy personnel entered the issue into the CAP asCR-JAF-2010-05591 .Analysis: The inspectors identified a performance deficiency because Entergypersonnel did not adequately plan and prevent unnecessary exposure during plannedwork activities. The finding was more than minor because it was associated with theRadiation Safety - Occupational Radiation Safety cornerstone attribute of program andprocess, and affected the cornerstone objective of protecting worker health and safetyfrom exposure to radiation. Specifically, the finding involved actual collective exposuregreater than five person-rem that was greater than 50 percent above the estimated orintended exposure. Additionally, this finding is similar to the "greater than minor"example provided in IMC 0612, Appendix E (Example 6.i, related to ALAM planning),The inspectors evaluated this finding in accordance with IMC 0609, Appendix C,"Occupational Radiation Safety Significance Determination Process," and determinedEnclosure

.218 that it was of very low safety significance (Green) because it involved an ALAMplanning issue and FitzPatrick's three year rolling average collective dose history wasless than 240 person-rem (146.593 person-rem for 2008-2010).This finding had a cross-cutting aspect in the area of Human Performance, work control,because Entergy personnel's planned work activities did not adequately incorporate thework site interferences or outage work coordination in the work control planning processtH,3(b)1.Enforcement: No violation of regulatory requirements was identified. The ALAM rule(10 CFR 20.1 101 ) Statements of Consideration indicates that compliance with theALARA requirement will be judged on whether the licensee has incorporated measuresto track and, if necessary, to reduce exposures, and not whether exposures and dosesrepresent an absolute minimum or whether the licensee has used all possible methodsto reduce exposures. The overall exposure performance of a nuclear power plant isused to determine its compliance with the ALAM rule. Entergy personnel entered theissue into the CAP as CR-JAF-2010-05591 . Since Fitzpatrick's three year rollingaverage collective dose (146.593 person-rem for 2008-2010) is below a three yearrolling average of 240 person-rem and FitzPatrick has an established ALARA program toreduce exposure consistent with the 10 CFR Part20.1101 Statement of Consideration,no violation of 10 CFR Part 20.1101(b) was identified. (FlN 05000333/2011005-02,Inadequate Work Planning for'A' Reactor Recirculation Pump Replacement)Failure to Follow Radiation Protection Procedureslntroduction: The inspectors identified a self-revealing NCV of very low safetysignificance (Green) of TS 5.4, "Procedures," because Entergy personnel did notadequately implement radiation protection procedures for completing RWP close outdocumentation. Specifically, Entergy staff did not complete the RWP close outdocumentation within 90 days after the R-19 refueling outage.Description: The R-19 refueling outage ended in November 2010. The inspectorsdetermined that, as of December 12, 2011, only 3 of 55 outage RWPs had close outdocumentation. Procedure EN-RP-105, "RadiologicalWork Permits," Revision 10,requires RWP close out and post job ALARA reviews to be completed within 90 daysfrom the end of the outage. By not completing the documentation within the 90 days,the inspector concluded that Entergy personnel could miss opportunities to identifylessons learned and implement corrective actions for improvement necessary forsubsequent outages. The inspectors noted that planning for R-20 outage had alreadybegun.Analvsis: The inspectors determined that failure to complete the RWP close outdocumentation within the 90 day requirement was a performance deficiency. Thefinding was more than minor because it was associated with the Radiation Safety -Occupational Radiation Safety cornerstone attribute of program and process andaffected the cornerstone objective of protecting worker health and safety from exposureto radiation. Specifically, Entergy staff did not complete RWP close out documentationto identify lessons learned and actions to reduce worker exposure in subsequentrefueling outages. Since planning for the R-20 outage had already begun, lessonslearned in the R-19 outage RWPs may not have been incorporated into the R-20 RWPsand potential additional unnecessary exposure could be avoided. Using IMC 0609,Enclosure

19Appendix C, "Occupational Radiation Safety Significance Determination Process," theinspectors determined that the finding screened as very low safety significance (Green)because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) asubstantial potentialfor overexposure, or (4) an impaired ability to assess dose.The finding had a cross-cutting aspect in the area of Human Performance, workpractices, because Entergy personnel did not communicate expectations regardingprocedural compliance tH.4(b)].Enforcement: FitzPatrick TS 5.4.1.a, requires that Entergy establish, implement, andmaintain procedures specified in Regulatory Guide (RG) 1.33, Revision 2, Appendix A,RG 1.33, Appendix A, Section 7.(e) specifies procedures for RWPs be established andimplemented, Procedure EN-RP-105 requires RWP close out and post job ALARAReviews to be completed within 90 days from the end of the outage. Contrary to theabove, as of December 1 2, 2011, only three of 55 RWP close out documents werecompleted for R-19 which ended in November 2010. Because this finding is of very lowsafety significance and has been entered into the licensee's CAP as CR-JAF-201 1-04152, this violation is being treated as an NCV, consistent with NRC EnforcementPolicy. (NCV 05000333/2011005-03, Failure to Follow Radiation ProtectionProcedures)2RS3 ln-Plant Airborne Radioactivitv Control and Mitiqation (71124.03 - 1 sample)a. lnspection ScopeDuring the period December 12 through 15, 2011 , the inspectors conducted thefollowing activities to verify that Entergy staff was controlling in-plant airborneconcentrations consistent with ALARA. lmplementation of these controls was reviewedagainst the criteria contained in 10 CFR Part20, applicable industry standards, andstation procedures.Inspection Planninoo The inspectors reviewed FitzPatrick's procedures for maintenance, inspection, anduse of respiratory protection equipment.o The inspectors verified there were no reported Pls.Use of Respiratorv Protection Deviceso The inspectors verified respiratory protection devices used were National Institute forOccupational Safety and Health (NIOSH) certified.e The inspectors verified the air used in SCBA was tested and met greater than orequalto Grade D quality.. The inspectors verified several individuals on the fire brigade and emergencyresponders were deemed fit to use the devices by a physician.r The inspectors verified training records for several individuals deemed fit to userespiratory devices.Enclosure 20Self-Contained Breathinq Apparatus for Emerqencv Use. The inspectors observed the monthly inspection of four SCBAs staged in the outagecommand center and the control room. The inspectors verified FitzPatrickpersonnel's capability to refill and transport bottles to and from the control room andthe operations support center during emergency conditions.o The inspectors verified control room operators and shift radiation protectiontechnicians were trained and qualified in the use of SCBAS. The inspectors alsoverified personnel assigned to fill bottles were trained and qualified to that task.r The inspectors verified appropriate mask sizes were available and that the controlroom operators on duty had no facial hair that would interfere with the sealingsurface of the face seal, The inspectors verified that corrective lenses for thoseoperators that require them were kept readily available in the control room.o The inspectors reviewed maintenance records for the four SCBAs inspected andverified any work performed was done by a contractor with certified training.b. FindinqsNo findings were identified.4.

4OA1 Performance Indicator Verification

.1 Mitiqatinq Svstems Performance lndex (5 samples)a. Lnspection ScopeThe inspectors reviewed FitzPatrick's submittal of the Mitigating Systems Performancelndex (MSPI) for the following systems for the period of October 1,2010 throughSeptember 30, 2011.. MSPI, emergency alternating current power systemo MSPI, high pressure injection systemr MSPI, heat removal systemo MSPI, residual heat removal systemr MSPI, cooling water systemsTo determine the accuracy of the performance indicator data reported during thoseperiods, the inspectors used definitions and guidance contained in NEI Document 99-02,"Regulatory Assessment Performance Indicator Guideline," Revision 6. The inspectorsalso reviewed station operator narrative logs, CRs, MSPI derivation reports, licenseeevent reports (LERs), and NRC integrated inspection reports to validate the accuracy ofthe submittals.b. FindinqsNo findings were identified.Enclosure

.222 Semi-Annual Trend ReviewInspection ScopeThe inspectors performed a semi-annual review of site issues, as required by lnspectionProcedure 71152, "Problem ldentification and Resolution," to identify trends that mightindicate the existence of more significant safety issues. In this review, the inspectorsincluded repetitive or closely-related issues that may have been documented by Entergypersonnel outside the CAP, such as trend reports, performance indicators, majorequipment problem lists, systems health reports, maintenance rule assessments, andmaintenance or CAP backlogs. The inspectors also reviewed the CAP database for thethird and fourth quarters of 2011 to assess CRs written in various subject areas(equipment problems, human performance issues, etc,), as well as individual issuesidentified during the NRC's daily CR review (Section 4OA2.1). The inspectors reviewedthe FitzPatrick quarterly trend report for the third quarter oI 2011, conducted under EN-LI-121, "Entergy Trending Process," to verify that Entergy personnel were appropriatelyevaluating and trending adverse conditions in accordance with applicable procedures.Findinos and ObservationsNo findings were identified.The inspectors evaluated a sample of departments that are required to provide input intothe quarterly trend reports, which included maintenance, engineering, and operationsdepartments. This review included a sample of issues and events that occurred over thecourse of the past two quarters to objectively determine whether issues wereappropriately considered or ruled as emerging, adverse, or monitored trends. Theinspectors verified that these issues were addressed within the scope of the CAP, orthrough department review and documentation in the quarterly trend report for overallassessment, For example, the inspectors noted that consistent with an increase inspurious upscale alarms in the average power range monitor (APRM) system that haveoccurred over several months, Entergy personnel had appropriately identified APRMlamp indications as an adverse trend and developed an action plan to address this issue.Annual Sample: Review of a Reactor Vessel Hioh Level Main Turbine Trip SwitchExceedino Surveillance Test Acceptance Criterialnspection ScopeThe inspectors selected CR-JAF-2010-06572 as a problem identification and resolutionsample for detailed review. This CR documented that on September 17,2010, the as-found trip setpoint for the 'A' reactor vessel high level main turbine trip level switch,06LS-121A, exceeded the TS acceptance criteria during a routine surveillance and wasrecalibrated. The inspectors reviewed the equipment failure and apparent causeevaluation (ACE). The inspectors assessed Entergy staff's problem identificationthreshold, cause analyses, extent of condition review, and the prioritization andtimeliness of their corrective actions to determine whether the staff was appropriatelyidentifying, characterizing, and correcting problems associated with this issue andwhether the planned and completed corrective actions were appropriate. The inspectorscompared the actions taken to the requirements of the CAP and 10 CFR 50,Appendix B.a.b..3a.Enclosure

b.23Findinqs and ObservationsNo findings were identified. The inspectors determined Entergy staff's overall responseto the issue was commensurate with its safety significance. The ACE and correctiveactions were reasonable and appropriate.Entergy personnel determined the apparent cause of the limit switch surveillance testfailure was related to an electrolytic capacitor installed across the input terminals of thelevel switch which either was leaking or could not respond to the input signal as fast asthe technician could input the signal. Therefore, the apparent causes documented werethat maintenance practices did not provide a consistent input rate of change into the limitswitch during testing, and, the capacitor was approximately 40 years old and was not inthe preventive maintenance (PM) program. Entergy staff also determined that, althoughthe specific component was out of tolerance high, the results of the entire loopcalibration showed that the function for this channel would have occurred within the TSrequired value. Entergy staff's extent of condition review identified additional capacitorsto be addressed. Corrective actions included revising the test procedure to eitherdevelop a new method to input the signal or direct that it be done slowly and to add theaffected capacitors to the PM program for replacement as soon as practicable.Entergy personnel determined the PM activities should be incorporated into the programand completed during the 2012 refueling outage. However, the inspectors identified theyhad been scheduled to be performed in the 2014 and 2016 refueling outages. Entergypersonnel initiated CR-JAF-2011-06470 and CR-JAF-2011-06472 to correct theconditions as intended in the original CR or reassess the priority in accordance with theCAP. This issue was determined to be minor because no equipment operability orfunctionality was significantly affected. In accordance with IMC 0612, "Power Reactorlnspection Reports," the above issue constituted a violation of minor significance that isnot subject to enforcement action in accordance with the Enforcement Policy.Annual Sample: Response to NRC lnformation Notice 2010-27. "Ve.ntilation Svstem',Inspection ScopeThe inspectors performed an in-depth review of condition report CR-JAF-2O11-00171,concerning the station review of NRC Information Notice (lN) 201 0-27 , "YentilationSystem Preventive Maintenance and Design lssues." This lN discusses recentoperating experience concerning ventilation system preventive maintenance and designissues, including instances involving the control room habitability system. Specifically, adesign weakness in the automatic ventilation shift logic at one nuclear power plantresulted in smoke from a fire outside of the plant being drawn into the control room whenthe system detected smoke and shifted control room ventilation into the smoke removalmode. At another plant, an earthquake caused a release of dust into the control room bythe ventilation system as the result of inadequate periodic cleaning and, along withnumerous alarms caused by the earthquake, contributed to the operators' decision tomanually scram the reactor. At a third plant, unexpected airflow rates with the standbyservice water pump house ventilation system operating in different modes led toidentification of a history of inadequate cleaning and maintenance on the intake screensand dampers..4a.Enclosure b.24The inspectors assessed Entergy's cause analysis, extent of condition reviews, and theprioritization and timeliness of corrective actions to determine whether Entergy staff wereappropriately identifying, characterizing, and correcting problems associated with this lNand whether the corrective actions were appropriate. The inspectors compared theactions taken to the requirements of the CAP and 10 CFR 50, Appendix B.Findinqs and ObservationsNo findings were identified.At FitzPatrick, there is a smoke detector located in the control room ventilation supplyducting, but it has no automatic functions associated with ventilation modes. Therefore,Entergy staff concluded that the first issue was not applicable. The inspectorsconsidered that this response was appropriate.In response to the second issue, Entergy staff noted that the control room ventilationsupply passes through two filters that are monitored by alarmed differential pressureswitches. However, dust that originates downstream the filters would have nothing tostop it from entering the control room. Therefore, Entergy staff assigned an action toperform a one-time inspection of the control room diffuser ducts to determine whetherperiodic cleaning was necessary. The actual action taKen by staff was that a visualinspection was performed from the outside of two of the 20 control room diffuser ducts,and noted to be free of dust. On this basis, Entergy personnel accepted the action ashaving been completed. The inspectors considered that a larger sampling and moreintrusive inspection would have provided a stronger basis for closure of this action.Concerning the third issue, Entergy staff noted there are two safety related ventilationfans that supply the screenwell, and that the associated intakes were inspectedincidentally during motor greasing that is performed every three years. The dampersassociated with these fans are replaced every 10 years and there is no periodicscheduled maintenance to lubricate the damper linkages because this replacementfrequency is considered adequate. Therefore, Entergy staff concluded that no correctiveactions were necessary with respect to this issue. The inspectors noted that, in the caseof the third issue, the lN indicated that "incidental" inspection had been ineffective inidentifying the need for ventilation system maintenance. Additionally, the inspectorsnoted that at least 18 CRs had been written in 2Q11 concerning safety and non-safetyrelated ventilation damper functional issues.The inspectors concluded that Entergy staff's review of lN 2010-27 was adequate, butpotentially could have identified more opportunities for preparedness and performanceimprovements. None of the observations made during the inspectors' review constitutedviolations of regulatory requirements.Follow-up of Events and Notices of Enforcement Discretion (71153 - 2 samples)(Closed) LER 05000333/201 1001-00 and -01 , Reactor Core lsolation Cooling SystemInoperable Longer than Allowed by Technical Specifications40A3.1Enclosure b.25lnspection ScopeOn January 7,2011, the RCIC system steam admission valve, 13MOV-131, did not fullyopen on demand while performing quarterly system surveillance testing. Entergy staff'stroubleshooting determined the cause to have been loose electrical connections in themotor operated valve (MOV) motor control circuit, apparently as a result or consequenceof maintenance that had been performed in September 2010, during R-19. A similarfailure of 13MOV-131 to fully open had occurred on October 29, 2010, however,following stem lubrication, the valve had operated properly. Given the intermittent natureof the actual cause of failure, Entergy staff concluded that 13MOV-131 should beconsidered to have been inoperable from the time that the RCIC system was required tobe operable during startup from R-19 (October 16, 2010) until the condition wascorrected on January 8, 2011. TS 3.5.3 requires the RCIC system to be operable inModes 1, 2, and 3, with reactor steam dome pressure greater than 150 pounds persquare inch gauge (psig) and provides an allowed outage time of 14 days.FindinqsIntroduction: The inspectors identified an NCV of very low safety significance (Green) of10 CFR 50, Appendix B, Criterion XVl, "Corrective Action," because Entergy staff did notpromptly correct the intermittent failure of RCIC steam admission valve 13MOV-131 tofully open on demand. Specifically, Entergy staff's troubleshooting performed inresponse to the October 29,2010 partial valve opening was of limited scope and notadequate to identify the cause of the intermittent failure at that time.Description: After the initial failure of 13MOV-131 to fully open on October 29, 2010,Entergy staff's troubleshooting identified that there was no grease evident on the anti-rotation yoke key way or the exposed portion of the valve stem. Following lubrication,the valve was successfully stroke tested and the quarterly RCIC system surveillance testwas completed satisfactorily. Therefore, Entergy personnel determined the cause of13MOV-131 not fully opening was due to excessive running load, caused by the lack oflubrication, which resulted in the torque switch opening.When 13MOV-131 did not fully open on January 7 , 2011, a more extensivetroubleshooting effort was undertaken by Entergy staff. This included static valvediagnostic testing and electrical inspections of the MOV and associated motor controlcircuit. As a result, the loose electrical connections in the MOV motor control circuitwere identified; specifically, loose coil connections for control relays 42-10 (open circuitDC contactor), 42-20 (open circuit seal-in contactor), and 42-2C (close circuit seal-incontactor) were identified. The loose connections were tightened and no otherdeficiencies were identified during the troubleshooting. No further issues were identifiedduring post-maintenance testing, which included dynamic valve diagnostic testing andperformance of the quarterly RCIC system surveillance test. The issue was entered intothe CAP as CR-JAF-2011-00123.Entergy staff concluded that the high resistance (loose) connection for open circuit seal-in contactor 42-20 was the cause of the intermittent failure of 13MOV-131 to fully open.When energized, current flow through the loose connection would cause its resistance toincrease due to heating, and the resultant decreasing current flow would eventuallycause the coil to drop out. Given this failure mechanism, Entergy staff concluded thatthe time for coil drop out to occur was repeatable; this was supported by the actualEnclosure 26opening times during the two partial valve opening occurrences. Since the RCIC pumpachieved rated discharge flow and pressure on both occasions that 13MOV-131 failed tofully open, the inspectors concluded that RCIC remained capable of performing itsdesign function during the period that this condition existed.The inspectors reviewed the original and revised LER, along with the associated CRs,apparent cause evaluations, and work documents. The inspectors concluded that testdata was consistent with Entergy staff's final postulated failure mechanism andconclusion that the RCIC system had remained capable of performing its design functionthroughout the period that the problem existed. However, the inspectors identified thatthe troubleshooting for the October 29,2Q10 valve failure focused primarily on the stemlubrication issue, rather than to thoroughly evaluate the overall mechanical and electricalcondition of 13MOV-131 , as would be appropriate for a critical component in a risksignificant system. As a result, Entergy personnel assumed that they had corrected theproblem when the valve was successfully stroked following stem lubrication. Theinspectors determined that, had the as-found characterization by Entergy personnelincluded static valve diagnostic testing (which, based on the final postulated failuremechanism, would have been completed satisfactorily), this erroneous conclusion wouldhave been avoided. The inspectors concluded that the inadequate scope of theOctober 29, 2010 troubleshooting plan for 13MOV-1 31 resulted in failure to identify theactual cause of the valve's incomplete opening problem. Consequently, the LimitingCondition for Operation (LCO) 3.5.3, 'RCIC System," was exceeded since the systemwas not restored to operable status within 14 days and the Unit was not placed in Mode3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and steam dome pressure reduced to less than or equal to 150 psigwithin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, prior to resolution of the problem.Analysis: The inspectors determined that the inadequate scope of troubleshootingperformed in response to the October 29, 2010 partial opening of 13MOV-131 was aperformance deficiency that was within FitzPatrick staff's ability to foresee and correct.The finding was more than minor because it affected the equipment performanceattribute of the Mitigating Systems cornerstone objective to ensure the reliability ofsystems that respond to initiating events to prevent undesirable consequences,Specifically, the loose electrical connections in the 13MOV"131 motor control circuitaffected the reliability of the RCIC system. The inspectors reviewed testing results andconcluded that the failure of 13MOV-131 did not prevent the RCIC pump from achievingrated discharge flow and pressure and the pump remained capable of performing itsdesign function during the period that the condition existed. The inspectors evaluatedthe finding using the Phase 1, "lnitial Screening and Characterization of Findings,"worksheet in Attachment 4 to IMC 0609, "Significance Determination Process." Theinspectors determined this finding was not a design qualification deficiency resulting in aloss of functionality or operability, did not represent an actual loss of safety function of asystem or train of equipment, and was not potentially risk significant due to a seismic,fire, or severe weather initiating event. Therefore, the inspectors determined the findingto be of very low safety significance (Green).The finding had a cross-cutting aspect in the area of Human Performance, work control,because Entergy personnel did not appropriately plan the 13MOV-131 troubleshootingactivity by incorporating consideration of the high risk significance of the RCIC systemlH.3(a)1.Enclosure

===.227

Enforcement:

10 CFR 50, Appendix B, Criterion XVl, "Corrective Action," states, in part,"Measures shall be established to assure that conditions adverse to quality, such asfailures, malfunctions, deficiencies, deviations, defective material and equipment, andnonconformance are promptly identified and corrected." Contrary to the above, followinga failure of the RCIC steam admission valve, 13MOV-131, to fully open on demand onOctober 29,2010, the cause of the equipment malfunction was not identified andcorrected by Entergy staff until after a second failure of the valve to fully open onJanuary 7,2011. The inadequate scope of troubleshooting performed by Entergy staff inresponse to the October 29, 2010 partial opening of 13MOV-131 also resulted in theLCO 3.5.3 being exceeded because the RCIC system was not restored to operablestatus within 14 days or be in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce steam dome pressureto less than or equal to 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Because this issue is of very lowsafety significance (Green) and Entergy personnel entered this issue into their CAP asCR-JAF-201 1-00123, this finding is being treated as an NCV consistent with the NRCEnforcement Policy. (NCV 05000333/2011005-04, Ineffective Gorrective Action forRCIC Steam Admission Valve Malfunction) This LER and its supplement are closed,(Closed) LER 05000333/201 1003-00, Safety Relief Valve Setpoints Outside ofAllowable TolerancesOn June 8, 2011, Entergy personnel determined the plant operated during the previousoperating cycle (Cycle 19) with less than nine operable safety relief valves (SRVs) asrequired by TS 3.4.3, "Safety/Relief Valves." TS 3.4.3 requires nine operable SRVswhen in Modes 1, 2 or 3. Entergy personnel had removed all 11 SRV pilot assembliesduring the previous R-19 and identified that five SRV pilot assemblies had as-found liftsetpoints outside the tolerance limits allowed by TS 3.4.3.1. Additionally, due to testequipment limitations, two SRV pilot assemblies could not be tested for set point driftdue to excessive pilot valve seat leakage. Entergy staff's root cause analyses forprevious SRV setpoint drift and pilot valve seat leakage issues determined that the mostprobable cause of the out of tolerance SRV setpoints was corrosion bonding betweenthe SRV pilot disc and seat, which has been an industry generic problem,Although Entergy staff has identified occurrences of SRV setpoint drift during eachrefueling outage since 2000, the inspectors determined that this most recent occurrencedid not constitute a violation of 10 CFR Part 50, Appendix B, Criterion XVl, for ineffectivecorrective action. The issue of 2-stage SRV setpoint drift due to pilot valve corrosionbonding has been a long standing industry generic problem, for which there is no singleidentified corrective action. Entergy staff has previously instituted a number ofrecommended strategies to correct the problem, such as installation of Stellite 21 pilotdiscs, installation of an electric lift system, and use of enhanced SRV insulation. Mostrecently (during R-19), the station implemented a phased replacement of 2-stage SRVswith 3-stage SRVs. The inspectors considered that this modification represented asubstantial corrective action that was implemented after the previous occurrence of SRVdrift in 2008, the effectiveness of which cannot yet be characterized.The failure of SRVs to operate within allowable tolerances describe in this LERconstituted a licensee-identified finding involving a violation of TS 3.4.3, "Safety ReliefValves." The enforcement aspects of the violation are discussed in Section 4OA7. ThisLER is closed.Enclosure===

284OAO Meetinos. Includino ExitExit Meetino SummarvThe inspectors presented the inspection results to Mr. M. Colomb and other members ofEntergy management at the conclusion of the inspection on January 23,2012. Theinspectors asked Entergy personnelwhether any materials examined during theinspection should be considered proprietary. No proprietary information was identifiedby Entergy personnel.4C.A7 Licensee-ldentified ViolationsThe following violation of very low safety significance (Green) was identified by thelicensee and is a violation of NRC requirements which meets the criteria of the NRCEnforcement Policy for being dispositioned as an NCV.. TS 3.4.3 requires that at least nine SRVs shall be operable in operating modes 1, 2,and 3. Contrary to this, on June 8,2011, Entergy personnel identified that the planthad operated in these modes during Cycle 19 with less than nine operable SRVs.Entergy personnel documented this condition in CR-JAF-2O1 1-0301 1 . Theinspectors determined this TS violation was of very low safety significance (Green)because it did not result in the loss of the overpressure relief safety function basedon operability of the electric lift system.ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Enterov PersonnelM, Colomb, Site Vice President

B. Sullivan, General Manager, Plant Operations

C. Adner, Manager, Operations

V. Bacanskas, Manager, Design Engineering

C. Brown, Manager, Quality Assurance, Entergy

R. Brown, Acting Manager, Radiation Protection

B. Finn, Director, Nuclear Safety Assurance

D. Koelbel, Sr. Engineer, Fire Protection

G. Sullivan, Acting Manager, Security

J. Pechacek, Manager, Licensing

D. Poulin, Manager, System Engineering

T. Raymond, Manager, Project Management

M. Reno, Manager, Maintenance

P. Scanlan, Manager, Programs and Components Engineering

M. Woodby, Director, Engineering

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened/Closed05000333/201 1 005-0105000333/201 1 005-02

05000333/201 1005-0305000333/201 1 005-04FINNCVNCVNCVMode Switch in ShutdownScram Function Inoperable in Excessof the TS Allowed Outage Time dueto Personnel Error (Section 1R13)Inadequate Work Planning for'A'Reactor Recirculation PumpReplacement (Section 2RS2)Failure to Follow RadiationProtection Procedures (Section2RS2)Ineffective Corrective Action forRCIC Steam Admission ValveMalfunction (Section 4OA3)

A-2

Closed

05000333i201 1001-00 and -01LERLERReactor Core lsolation CoolingSystem Inoperable Longer thanAllowed by Technical Specifications(Section 4OA3)Safety Relief Valve SetpointsOutside of Allowable Tolerances(Section 4OA3)050003331201 1003-00DiscussedNone

Section 1R01: Adverse Weather ProtectionProcedures:AOP-13, "High Winds, Hurricanes and Tornadoes," Revision 13AP-12.04, "Seasonal Weather Preparations," Revision 1 8OP-s1A, "Reactor Building Ventilation and Cooling System," Revision 49Documents:DBD-092, "Design Basis Document for the

EDG Building Heating and Ventilation System,"Revision 6Condition Reports:CR-JAF-2O1 1-05652

Section 1R04: Equipment AliqnmentProcedures:OP-19, "Reactor Core lsolation Cooling System," Revision 48OP-21, "Emergency Service Water," Revision 37OP-22, "Diesel Generator Emergency Power", Revision 57Condition ReportscR-JAF-2011-04768cR-JAF-2o11-04770CR-JAF-201

1-05159

Section 1R05: Fire ProtectionProcedures:OP-33, "Fire Protection," Revision 54PFP-PWR04, "Battery Room Complex/Elev 272',282' Fire ArealZone lll/BR-1,

BR-2, lV/BR-3,BR-4, XVI/BR-S," Revision 2PFP-PWR32,"Emergency Diesel Generator Spaces-North/Elev. 272'Fire ArealZone Vl/EG-3,EG-4.

EG-6," Revision 3Attachment

A-3PFP-PWR11, "Cable Spreading Room / Elev. 272'Fire ArealZone Vll/CS-1," Revision 2PFP-PWR27, "Reactor Building / Elev. 344' Fire ArealZone l)URB-1A," Revision 4PFP-PWR28, "Reactor Building / Elev. 369' Fire ArealZone

IXRB-1A," Revision 7Documents:JAF-RPT-O4-00478, "JAF Fire Hazards Analysis," Revision 2Condition Reports:cR-JAF-2o11-05872

Section 1RO7: Heat Sink PerformanceProcedures:EN-DC-316, "Heat Exchanger Performance and Condition Monitoring," Revision 3ENN-SEP-HX-007, "JAF Heat Exchanger Program," Revision 0SEP-SW-0O1, "FitzPatrick

NRC Generic Letter 89-13 Service Water Program," Revision 1Documents:AP-19.12 Service Water lnspection data sheet for October 26,2011 inspection of 'B' EDGjacket water coolerEN-DC-316 Heat Exchanger lnspection Data Sheet for October 26,2011 inspection of 'B' EDGjacket water coolerProgram Health Report, Heat Exchanger Program, third quarter 2011

Section 1R11: Licensed Operator Requalification ProqramProcedures:AOP-1, "Reactor Scram," Revision 44AOP-3, "High Activity in Reactor Coolant or Off-Gas," Revision 16AOP-4, "Explosion in Air Ejector Discharge Piping," Revision 5AOP-32, "Unplanned Power Change," Revision 11AOP-40, "Main Steam Line Break," Revision 10AOP-62, "Loss of Feedwater Heating," Revision 10EOP-?, "RPV Control," Revision 9EOP-5, "Secondary Containment Control," Revision 8EOP-6, "Radioactivity Release Control," Revision 8Section 1R12: Maintenance EffectivenessProcedures:EN-DC-203, "Maintenance Rule Program," Revision 1EN-DC-204, "Maintenance Rule Scope and Basis," Revision 2EN-DC-205, "Maintenance Rule Monitoring," Revision 3OP-37, "Containment Atmosphere Dilution System," Revision 79Documents:JAF-RPT-CAD-02312, "Maintenance Rule Basis Document System 27, Primary ContainmentAtmosphere Control and Dilution," Revision 11JENG-1

1-0041, "Maintenance Rule (aX1) Action Plan, Containment Air Dilution," Revision 0JENG-1

1-0041, "Maintenance Rule (aX1) Action Plan, Containment Air Dilution," Revision 1Attachment

A-4Maintenance Rule Quarterly Report, third quarter 2011System Health Report, CAD, third quarter 2011System Health Report, EDG, third and second quarter 2011DBD-093, "Design Basis Document for the Emergency Diesel Generator (EDG)," Revision 11JAF-RPT-EDG-02303, "Maintenance Rule Basis Document System 93, Emergency DieselGenerator," Revision 9JAF-RPT-DGV-02301, "Maintenance Rule Basis Document System 92, Emergency DieselGenerator Ventilation," Revision 5Condition Reports:cR-JAF-2o10-00298cR-JAF-2010-01712CR-JAF-2010-03871CR-JAF-2011-02584oR-JAF-201

1-03964oR-JAF-2010-00014CR-JAF-2010-00310CR-JAF-2010-00320Work Orders:CR-JAF-2O10-00739CR-JAF-2010-01270CR-JAF-2o10-03525CR-JAF-2010-04660cR-JAF-2o10-05965cR-JAF-z010-08533CR-JAF-201

1-00667CR-JAF-2O1 1-00689CR-JAF-2O11-01943CR-JAF-2011-02443CR-JAF-2011-02733CR-JAF-2011-02770CR-JAF-2011-02834cR-JAF-zo11-02973cR-JAF-2o11-04873cR-JAF-2o11-04945285832

Section 1R13: Maintenance Risk Assessments and Emersent Work ControlProcedures:AP-05.13, "Maintenance During

LCOs," Revision 10AP-10.10, "On-Line Risk Assessment," Revision 7EN-WM-104, "On Line Risk Assessment," Revision 6

Section 1R15: Operabilitv Determinations an{ Functionalitv AssessmentsProcedures:EN-LI-102, "Corrective Action Process," Revision 17EN-OP-1 04, "Operability Determination Process," Revision 5OP-438, "24VDC Power System," Revision 7ST-2XA, "RHR Service Water Loop A Quarterly Operability Test (lST)

ST-2XA," Revision 13Documents:DBD-071 Tab lll, "Design Basis Document for the Electrical Distribution System 125V andTurbine building rounds data December 10-16,2011

Section 1R18: Plant ModificationsProcedures:EN-DC-136, "Temporary Modifications," Revision 6OP-50, "Equipment and Floor Drain System," Revision 31ARP-09-4-1-20, "Rx Bldg Equip Sump B Temp Hi," Revision 2Attachment

A-5Documents:EC 30962, "Temporary Alarm Setpoint Change for 20TlS-5348 Reactor Building EquipmentSump Temp"Condition Reports:CR-JAF-2O1 1-05676cR-JAF-2011-05677

Section 1R19: Post Maintenance TestinqProcedures:ISP-91-1(10600), "10600 Bus 4 kV Emergency Bus Degraded Voltage Timer InstrumentCalibration," Revision 4ST-24J, "RCIC Flow Rate and Inservice Test (lST)," Revision 41ISP-19-2A, "Post Accident Off-Gas (Stack) High Range Radiation Monitor A FunctionalTesUCalibration," Revision 2ST-9BA, "EDG A and C Full Load Test and

ESW Pump Operability Test," Revision 12ESP-22.004, "EDG B & D Fuel OilTransfer Pump Operational Check," Revision 0Condition Reports:CR-JAF-2011-06244CR-JAF-201 1-05805

Section 2RS2: Occupational

ALARA Planninq and ControlsProcedures:EN-RP-105, "Radiological Work Permits," Revision 10EN-RP-110, "ALARA Program," Revision 8EN-RP-1 10-01, "ALARA Initiative Deferrals," Revision 0EN-RP-110-02, "Elemental Cobalt Sampling," Revision 0EN-RP-110-03, "Collective Radiation Exposure (CRE) Reduction Guidelines," Revision 0EN-RP-110-04, "Radiation Protection Risk Assessment Process," Revision 1EN-RP-110-05, "ALAM Planning and Controls," Revision 0EN-RP-110-06, "Outage Dose Estimating and Tracking," Revision 09ondition Reports:CR-JAF-2010-02871CR-JAF-2010-03770CR-JAF-2010-03884CR-JAF-2010-04833CR-JAF-2010-05591CR-JAF-2010-06018CR-JAF-2010-06064CR-JAF-2010-06712cR-JAF-zo10-06909cR-JAF-2o11-04152

==Section 2RS3: In-Plant Airborne Radioactivitv Control and MitisationSCBAs:Case Requlator lD120

214241

215718

2209104 2188Attachment==

A-6Section 4OA1 : Performance Indicator VerificationDocuments:JAF-RPT-05-00047, "Mitigating System Performance lndex (MSPI) Basis Document,"Revision 3Condition Reports:oR-JAF-2011-05721

Section 4OA2: ldentification and Resolution of ProblemsProcedures:EN-Ll-121, "Entergy Trending Process," Revision 10EN-DC-143, "Engineering Health Reports," Revision 13EN-DC-159, "System Monitoring Program," Revision 6EN-Ll-102, "Corrective Action Process," Revision 17EN-OP-111, "Operational Decision Making lssue (ODMI) Process," Revision 6EN-OP-1 15, "Conduct of Operations," Revision 12EN-OP-1 15-07, "Component Deviations," Revision 0ISP-3-10, "Feedwater Control System - High Reactor Level Pump Trip Functional andCalibration," Revision 1 9Documents:James A. FitzPatrick Quarterly Trend Report, third quarter 2011Open Operational Decision Making lssues for 2011Closed Operational Decision Making lssues for 2011System Health Report, Recirculation System, second and third quarter 2011System Health Report, Neutron Monitoring, second and third quarter 2011Performance Summary System Engineering, November 2010 - November 2Q11Performance Summary Design Engineering, November 2010 - November 2011Performance Summary Operations, November 2010 - November 2011Performance Summary Maintenance, November 2010 - November 2011PMCR 10199, "Perform

PMCR for Capacitors"JAFLO-2O11-0012, Snapshot Assessment - System Health/Monitoring, March - April 2011JAFLO-2O11-0068, Focused Self-Assessment - Conduct of MaintenanceJAFLO-2O11-0070, Focused Self-Assessment - Weakness in Operator FundamentalsJAFLO-2O1

1-0002 Quarterly Trend Report Department SubmittalsCondition Reports:cR-JAF-2010-06572CR-JAF-2011-01037CR-JAF-201 1-01 166CR-JAF-201

1-01356cR-JAF-2011-02316CR-JAF,2o1

1-03588CR-JAF-201

1-03820Work Orders:5233286352332864CR-JAF-2011-06470CR-JAF-2011-04043cR-JAF-2011-04341CR-JAF-2011-05125CR-JAF-2O1

1-05130CR-JAF-2011-05241CR-JAF-2o11-052425233286552332866cR-JAF-2O11-06472cR-JAF-2o11-05243CR-JAF-2011-05245cR-JAF-2011-05246CR-JAF-201

1-05973CR-JAF-201

1-06051CR-JAF-2o1

1-0609552332867Attachment

CFRACEADAMSALARAAPRMARPCAPCRDBDECEDGEntergyFitzPatrickHPCItMcINISTKVLCOLERMGMOVMSPINCVNEINIOSHNRCODCMPARSPIPMPMTpsigRRBRCrCRETSRGRHRRPSRWPA-7

LIST OF ACRONYMS

Title 10 of the Code of Federal Regulationsapparent cause evaluationAgencywide Documents Access and Management Systemas low as reasonably achievableaverage power range monitoralarm response procedurecorrective action programcondition reportdesign basis documentengineering changeemergency diesel generatorEntergy Nuclear NortheastJames

A. [[FitzPatrick Nuclear Power Planthigh pressure coolant injectioninspection manual chapterinformation noticein-service testingkilovoltlimiting condition for operationlicensee event reportmotor-generatormotor operated valvemitigating systems performance indexnon-cited violationNuclear Energy InstituteNational Institute for Occupational Safety and HealthNuclear Regulatory CommissionOffsite Dose Calculation ManualPublicly Available Recordsperformance indicatorpreventive maintenancepost-maintenance testi ngpounds per square inch gaugerefueling outagereactor buildingreactor core isolation coolingrad iological effluent occurrencesregulatory guideresidual heat removalreactor protection systemradiation work permitAttachment]]

SCBASD [[]]

PSRVSS [[CsSTSWTSTSCUFSARVDCWOA-8self-contained breathing apparatussignificance determination processsafety relief valvestructures, systems, or componentssurveillance testswitchtechnical specificationtechnical support centerupdated final safety analysis reportvolt direct currentwork orderAttachment]]