IR 05000271/2011002

lR 05000271-11-002;01/01/2011 - 03/31/2011; Vermont Yankee Nuclear Power Station; Post-Maintenance Testing; Event Follow-up

ML111190386
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	04/29/2011
From:	Jackson D E NRC/RGN-I/DRP/PB5
To:	Michael Colomb Entergy Nuclear Operations
Jackson, D E RI/DRP/PB5/610-337-5306
References
IR-11-002
Download: ML111190386 (34)

Text

ApriT 29, 2OLLMr. Michael ColombSite Vice PresidentEntergy Nuclear Operations, Inc.Vermont Yankee Nuclear Power StationVernon. W 05354

SUBJECT: VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATEDI NSPECTION REPORT 0500027 1 1201 1002

Dear Mr. Colomb:

On March 31,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your Vermont Yankee Nuclear Power Station. The enclosed inspection report documents theinspection results, which were discussed on April 1 1,2011, with you and other members of yourstatf.The inspection examined activities performed under your license as they relate to safety andcompliance with the Commission's rules and regulations, and with the conditions of yourlicense. The inspectors reviewed selected procedures and records, observed activities, andinterviewed personnel.This report documents two self-revealing findings of very low safety significance (Green).These iindings were determined to involve violations of NRC requirements. However, becauseof the very low safety significance and because they have been entered into your correctiveaction program (CAP), the NRC is treating these findings as non-cited violations (NCV),consisient with Section 2.3.2.a of the NRC's Enforcement Policy. lf you contest any NCV, youshould provide a response within 30 days of the date of this inspection report, with the basis foryour denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,Washington DC 20555-0001; witfr copies to the RegionalAdministrator, Region l; the Director,Office oi Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001: and the NRC Senior Resident Inspector at Vermont Yankee. In addition, if you disagreewith any cross-cutting aspects assigned to the findings in this report, you should provide aresponie 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 lnspectorat Vermont Yankee.

M. ColombIn 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 (ADAMS). ADAMS is accessible from the NRC Web Site athttp://www,nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).Donald E. JacksoProjects Branch 5Division of Reactor Projects50-271DPR-28Inspection Report No. 0500027 1 12011002M

Attachment:

Supplemental InformationDistribution via ListServ

Sincerely,&ADocket No.License No.

Enclosure:

cc Mencl: 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 (ADAMS). ADAMS is accessible from the NRC Web Site athttp://www.nrc.gov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Donald E. Jackson, ChiefProjects Branch 5Division of Reactor ProjectsDocket No. 50-271License No. DPR-28

Enclosure:

lnspection Report No. 0500027 1 1201 1002w/

Attachment:

Supplemental InformationDistribution w/encl: (via E-mail)W. Dean. RAD. Lew, DRAD. Roberts. DRPJ. Clifford, DRPC. Miller, DRSP. Wilson, DRSS. Bush-Goddard, OEDOD. Jackson. DRPD. Spindler, DRP, SRI(RlORAMAILResource)(RlORAMAlLResource)(Rl DRPMAlLResource)(Rl DRPMAlLResource)(Rl DRSMAlLResource)(Rl DRSMAlLResource)A. Ziedonis, DRP, Acting SRIT. Setzer. DRPD. Dodson. DRPS. Rich, DRP, RlA. Rancourt, DRP, OARidsNrrPMVermontYankee ResourceRidsNrrDorlLl 1 -1 ResourceRO PreportsResource@n rc.qovSUNSI Review Complete: TCS (Reviewer's Initials)DOCUMENT NAME: G:\DRP\BRANCHS\Reports\DraftsVY 201 1 002.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" = Copy with attachmenUenclosure "N" = No copyM1111190386OFFICERI/DRPlhpRI/DRPRI/DRPRI/DRPNAMEAZiedonis/tcs forTSetzer/tcsDJackson./dejDATE04t27t1104t27t1104t29t11 Docket No.:License No.:Report No.:Licensee:Facility:Location:Dates:Inspectors:Approved by:1U.S. NUCLEAR REGULATORY COMMISSIONREGION I50-271DPR-280500027112011002Entergy Nuclear Operations, Inc.Vermont Yankee Nuclear Power StationVernon, Vermont 05354-9766January 1,2011 through March 31,2011D. Spindler, sr. Resident lnspector, Division of Reactor Projects (DRP)S. Rich, Resident InsPector, DRPDonald E. Jackson, ChiefProjects Branch 5Division of Reactor ProjectsEnclosure 2

SUMMARY OF FINDINGS

lR 0500027112011002;0110112011 - 0313112011; Vermont Yankee Nuclear Power Station;Post-Maintenance Testing; Event Follow-up.This report covered a three-month period of inspection by resident inspector staff and region-based inspectors. Two Green, self-revealing findings, which were determined to be non-citedviolations (NCV), were identified. The significance of most findings is indicated by their color(Green, White, Yellow, Red) using lnspection Manual Chapter (lMC) 0609, "SignificanceDetermination Process." The cross-cutting aspects for the findings were determined using IMC0310, "Components Within The Cross-Cutting Areas." Findings for which the significancedetermination process does not apply may be Green or be assigned a severity level after NRCmanagement review. The NRC's program for overseeing the safe operation of commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4,dated December 2006.

Cornerstone: Mitigating Systems.

Green.

A self-revealing, non-cited violation (NCV) of very low safety significance (Green) ofTechnical Specifications 6.4, "Procedures," was identified for inadequate implementation ofEntergy procedure EN-MA-118, "Foreign Material Exclusion," Revision 6, which resulted inforeign material intrusion into the Residual Heat Removal Service Water (RHRSW) system.Specifically, Entergy did not establish a Foreign Material Exclusion (FME) Zone l around theopen RHRSW system between completing the closeout inspection and system closurefollowing pump replacement. Entergy's immediate corrective actions included conducting a"stand down," reinforcing the standards and requirements for FME controls and generalprocedural compliance, as well as reinforcing expectations for the attention to detail of workpractices. Entergy entered the issue into their corrective action program to evaluate foradditional corrective measures.The inspectors determined that the finding was more than minor because it was associatedwith the Equipment Performance attribute of the Mitigating Systems cornerstone, andaffected the cornerstone objective to ensure the availability of systems that respond toinitiating events to prevent undesirable consequences, (i.e., core damage). Specifically,foreign material made its way into the'A'Residual Heat Removal Heat Exchanger(RHR HX) and rendered the'A' RHRSW train inoperable for several days. A review of NRClnspection Manual Chapter (lMC) 0612, Appendix E, "Minor Examples," revealed that nominor examples were applicable to this finding. The inspectors used IMC 0609.04, "Phase 1- Initial Screening and Characterization of Findings," and determined that the findingrequired a Phase 2 review because the 'A' RHRSW train had an actual loss of safetyfunction for greater than its allowed outage time (7 days). This finding was assessed usingIMC 0609 and was determined to be of very low safety significance (Green) based on aPhase 2 analysis. The finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Practices component, because Entergy personnel did not follow EN-MA-1 18. Specifically, they did not establish a FME Zone 1 after the system closeout inspection.tH.4(b)l (Section 1 R1 e)

4.

Green.

A self-revealing, Green NCV of Technical Specification 6.4, "Procedures," wasidentified in which maintenance and planning personnel did not involve engineeringpersonnel as required by Entergy procedure EN-MA-101 , "Fundamentals of Maintenance,"Revision 9, and EN-WM-105, "Planning," Revision 8, resulting in the incorrect material beingused to replace the gasket on the flange of High Pressure Coolant Injection System (HPCI)steam trap 23T-3. Entergy ultimately replaced the gasket with the correct material andentered this issue into their corrective action program.The inspectors determined that the finding was more than minor because it was associatedwith the Human Performance attribute of the Mitigating Systems cornerstone, and affectedthe cornerstone objective to ensure the availability of systems that respond to initiatingevents to prevent undesirable consequences. The finding was determined to be of very lowsafety significance (Green) in accordance with IMC 0609, Appendix A, "Determining theSignificance of Reactor Inspection Findings for At-Power Situations," using SignificanceDetermination Process (SDP) Phases 1,2 and 3. A Region I Senior Reactor Analyst (SRA)conducted a Phase 3 analysis because the Phase 2 analysis indicated that the finding hadthe potential to be greater than very low safety significance (Greater than Green). Thisfinding had a cross-cutting aspect in the Human Performance cross-cutting area, DecisionMaking component, because Vermont Yankee personnel did not obtain interdisciplinaryinput on the decision to use a different, incorrect gasket material in a steam trap in the HPCIsystem. H.1(a) (Section 4OA3)

Other Findings

Violations of very low safety significance that were identified by the licensee have beenreviewed by the inspectors. Corrective actions taken or planned by the licensee have beenentered into the licensee's corrective action program. These violations and corrective actiontracking numbers are listed in Section 4OA7 of this report.Enclosure

.15

REPORT DETAILS

Summarv of Plant StatusVermont Yankee (W) Nuclear Power Station began the inspection period operating at 100percent power. On February 14,2011, W performed a planned power reduction to 58 percentpower to perform main steam line isolation valve testing, main turbine stop valve testing, and arod pattern adjustment. W returned to 100 percent power on February 15,2011, and remainedat or near 100 percent power for the duration of the inspection period.1. REACTORSAFETYCornerstones: Initiating Events, M itigating Systems, Barrier Integrity1R01 Adverse Weather Protection (71111.01)lFpendinq Adverse Weatherlnspection Scope (1 sample)The inspectors reviewed Entergy's procedures in order to evaluate the process forimplementation of extreme cold temperature preparedness. This review was conductedfrom January 21, 2011, through January 24,2011, due to forecasted overnight lowtemperatures below negative 15 degrees Fahrenheit. The inspectors reviewed adverseweather information contained in Vermont Yankee's lndividual Plant Examination forExternal Events and compared it to the actions specified in Entergy operating procedure(OP) 3127, "Natural Phenomena," Revision 26 and OP 2196, "Seasonal Preparedness,"Revision 31. The inspectors reviewed documents, interyiewed personnel and performeda walkdown of the reactor building, turbine building and intake structure to verify thatactions required by the above procedures had been taken and that indoor temperatureswere not low enough to impact equipment operability.FindinqsNo findings were identified.External Floodino ReadinessInspection Scope (1 sample)The inspectors reviewed Entergy's flood protection barriers and procedures for copingwith externalflooding. The inspectors reviewed externalflooding information containedin the Updated Final Safety Analysis Report (UFSAR) and lndividual Plant Examinationfor External Events, and compared it to the actions specified in OP 3127, "NaturalPhenomena," Revision 26. The inspectors performed walkdowns of the switchgearrooms, cooling towers, intake structure, and outside areas. They also examined theEnclosurea.b.a.,2

===.16 equipment specified in the OP (sump pumps, floor drain plugs, sandbags, etc.) todetermine if it was available for use. The inspectors also reviewed a sample of externalflooding-related conditions identified in W's CAP to determine if they were appropriatelyidentified and corrected. The documents reviewed are listed in the Attachment.b. FindinqsNo findings were identified.

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a.7also assessed housekeeping and component labeling. ln addition, the inspectorsreviewed the system health reports, and evaluated a sample of previously identifieddeficiencies to determine if they had been properly addressed. The inspectorsperformed a search of the corrective action program for equipment alignment problemsto verify that Entergy was identifying problems at an appropriate threshold and resolvingthem appropriately. These activities constituted one complete equipment alignmentinspection sample. Documents reviewed are listed in the Attachment.b. FindinqsNo findings were identified.

1R05 Fire Protection (71111.05)Quarterlv lnspection (7 1 1 1 1

===.05 O)Inspection Scope (5 samples)The inspectors performed inspections of five fire areas based on a review of theVermont Yankee Safe Shutdown Capability Analysis and the Fire Hazards

Analysis.

Theinspectors reviewed Entergy's fire protection program to determine the specified fireprotection design features, fire area boundaries, and combustible loading requirementsfor the selected areas. The inspectors verified, consistent with applicable administrativeprocedures, that combustibles and ignition sources were adequately controlled; passivefire barriers, manualfire-fighting equipment, and detection and suppression equipmentwere appropriately maintained; and compensatory measures for out-of-service,degraded, or inoperable fire protection equipment were implemented in accordance withEntergy's fire protection program. The inspectors evaluated the fire protection programfor conformance with the requirements of License Condition 3.F. The documentsreviewed are listed in the Attachment. The following fire areas were inspected:. Turbine Lube OilTank and Storage Room, FZ-6;. Control Building E\.262'Cable Vault, FA ASD, FZ-2;. HPCI Room, FZRB-2;. 'B'EDG Room with Barrier Breach, FA-9; and. Main, Auxiliary and Startup Transformers.FindinosNo findings were identified.b.Enclosure===

81R06 Flood Protection Measures (71111.06 - 1 sample)b. FindinqsNo findings were identified.1R1 1 Licensed Operator Requalification Proqram (71111.11)lnternal FloodinoInspection ScopeThe inspectors reviewed Entergy's flood protection design and barriers for coping withinternalflooding on the Reactor Building 252' elevation. The inspectors reviewedinternalflooding information contained in Vermont Yankee's lndividual Plant Examinationfor External Events (IPEEE) and the internalflooding design basis document. Theinspectors performed a walkdown of the area to ensure equipment and structuresneeded to mitigate an internalflooding event were as described in the IPEEE and thedesign basis document. Additionally, the inspectors reviewed CRs related to internalflooding to ensure identified problems were properly addressed for resolution.Documents reviewed are listed in the Attachment. These activities constituted oneinternal flood protection measures inspection sample.b.Quarterlv Inspection (71111.1 1O)Inspection Scope (1 sample)The inspectors observed a simulator-based licensed operator requalification (LOR)exam on February 7,2011. The inspectors assessed the performance of risk significantoperator actions, including the use of emergency operating procedures. The inspectorsevaluated crew performance in the areas of clarity and formality of communications;ability to take timely actions; prioritization, interpretation, and verification of alarms;procedure usage; control board manipulations; and command and control. Theinspectors also compared the simulator configuration with the actual control boardconfiguration. Finally, the inspectors verified that evaluators were identifying anddocumenting crew performance problems. The documents reviewed are listed in theAttachment.FindinqsNo findings were identified.Enclosure

1R12 Maintenance Effectiveness (7111 1.12)Quarterly Inspection (7 1 1 1 1

.124 )a. Inspection Scope (3 samples)The inspectors reviewed performance-based problems involving selected in-scopestructures, systems and components (SSCs) to assess the effectiveness of themaintenance program. The reviews focused on the following aspects when applicable:. Proper Maintenance Rule scoping in accordance with 10 CFR 50.65;o Characterization of reliability issues;. Charging system and component unavailability;. 10 CFR 50.65 paragraph (aX1) and (a)(2) classifications;. ldentifying and addressing common cause failures;. Appropriateness of performance criteria for SSCs classified paragraph (aX2); anO. Adequacy of goals and corrective actions for SSCs classified paragraph (aX1).The inspectors reviewed the applicable system health reports, maintenance backlogs,and Maintenance Rule basis documents. The documents reviewed are listed in theAttachment. The following structures, systems and components were inspected:. Augmented Off-gas System;. Instrument Air System; ando Service Air System.b. FindinqsNo findings were identified.1 R13 Maintenance Risk Assessments and Emeroent Work Control (71111 .13)a. Inspection Scope (5 samples)The inspectors evaluated five maintenance risk assessments for planned and emergentmaintenance activities to verify that the appropriate risk assessments were performedprior to removing equipment for work. The inspectors reviewed maintenance riskevaluations, maintenance plans, work schedules, and control room logs to determine ifconcurrent or emergent maintenance or surveillance activities significantly increased theplant risk. The inspectors reviewed risk assessments to determine if they wereperformed as required by 10 CFR 50.65 paragraph (aX4) and implemented inaccordance with Entergy's administrative procedure (AP) 0172, "Work Schedule RiskManagement - Online." When emergent work was performed, the inspectors observedactivities to determine if plant risk was promptly reassessed and managed. Theinspectors conducted plant walkdowns to verify that appropriate risk managementEnclosure

10actions had been taken. The documents reviewed are listed in the Attachment. Thefollowing maintenance activities were inspected:. Work Week 1101 - Emergent Work on 'A' RHRSW and RHR Trains;. Work Week 1103 -'B' Diesel Generator Testing and Battery B-AS-2 Maintenance;. Work Week 1105 - Service Water Valve testing;. Work Week 1107 - Emergent Work on HPCI; and. Work Week 1111 - Service Water Strainer maintenance and Standby Liquid ControlSurveillance.b. FindinqsSee Section

==4OA7 .

1R15 Operabilitv Evaluations

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a. Inspection Scope

(5 samples)The inspectors reviewed five operability evaluations associated with degraded or non-conforming conditions to assess the acceptability of the evaluations, the use and controlof applicable compensatory measures, and compliance with Technical Specifications.The inspectors reviewed and compared the technical adequacy of the evaluations withthe Technical Specifications, UFSAR, associated design basis documents, andEntergy's procedure EN-OP-104, "Operability Determinations." The documentsreviewed are listed in the Attachment. The inspectors reviewed evaluations of thefollowing degraded or non-conforming conditions:. CR 2011-00301 -'B' RHRSW Pump Met In-service Testing Action Limit for LowPump Differential Pressure;. CR 2011-00694 - Main Diesel Fuel Oil Flash Point at Procedural Lower Limit;. CR 2011-00876 and 2011-00880 - Water Leakage Found on Cylinder AdapterPlates on 'B' Emergency Diesel Generator (DG-1-B) ;. CR 2011-00773 - RCIC Environmental Qualification (EQ); anO. CR-2010-0556, 2010-05023,2011-00193, 2011-00652, and 2011-00713 - GeneralElectric Hitachi Design Life of 'D' and 'S' Lattice Marathon Control Rod Blades.b. FindinosNo findings were identified.Enclosure 111R18 Plant Modifications (71111.18)Permanent Plant Modificationsa. Inspection Scope (2 samples)The inspectors reviewed EC21288, "Replace V76-38 with a New Check Valve," andEClT444, "ChemicalTreatment Connections to the Spent Fuel Cooling (SFPC)System," to ensure that they did not adversely affect the availability, reliability, orfunctional capability of any risk-significant SSCs. The inspectors reviewed theengineering change packages, and observed the systems in operation following theimplementation of the modifications. The documents reviewed are listed in theAttachment.b. FindinqsNo findings were identified.

1R19 Post-Maintenance Testinq (71111.19)Inspection Scooe (7 samples)The inspectors reviewed seven post-maintenance test (PMT) activities on risk-significantsystems. The inspectors reviewed these activities to determine whether test acceptancecriteria were clear and consistent with design basis documents. When testing wasdirectly observed, the inspectors determined whether installed test equipment wasappropriate and controlled, and whether the test was performed in accordance with10 CFR Part 50, Appendix B, Criterion Xl, "Test Control," and applicable stationprocedures. Upon completion, the inspectors performed a walkdown to verify thatequipment was returned to the proper alignment necessary to perform its safety function,and evaluated whether conditions adverse to quality were entered into the CAP forresolution. The documents reviewed are listed in the Attachment. The inspectorsreviewed the PMTs performed for the following maintenance activities:r RHR Pumps'A'and 'C'and RHR Service Water Pump 'A'Testing Following RHRHeat Exchanger Work;e Fire Protection Check Valve V76-3B Replacement;o 'B' Service Water Pump Replacement;. 'B' Emergency Diesel Generator Overhaul;o Repair of HPCI Steam Trap 23T-3;. 'C'Circulating Water Pump Replacement; and. 'B'RHR Service Water Pump Replacement.Enclosure

b.12FindinqsIntroduction: A self-revealing, NCV of very low safety significance (Green) of TechnicalSpecifications 6.4, "Procedures," was identified for inadequate implementation ofEntergy procedure EN-MA-1 18, "Foreign Material Exclusion," Revision 6, which resultedin foreign material intrusion into the RHRSW system. Specifically, Entergy did notestabfish a procedurally required FME Zone l around the open RHRSW systembetween completing the closeout inspection and system closure following pumpreplacement.Discussion: On December 27,2010, Entergy began removal of the 'C' RHRSW pumpfor a planned replacement. During the planned replacement of the 'C' RHRSW pump,the 'A'train of RHRSW was planned to remain in an operable status, since the'A'RHRSW pump was not planned to be affected by the 'C' pump replacement, and sinceone RHRSW pump provides sufficient capacity to perform the safety function of the 'A'RHRSW train. During the work activity, the area was controlled as a FME Zone 2, whichrequires some FME boundaries and work practices, but does not require materialentering the zone to be either tracked on a log or tied down as is required in a FME Zone1 . On December 30, 2010, Entergy personnel performed a closeout inspection of the 'C'RHRSW pump and piping prior to final pump assembly, but did not upgrade the area toa FME Zone 1. EN-MA-118, "Foreign Material Exclusion," states that a FME Zone 1should be established, "when a final visual inspection of internal cleanliness beforesystem closure is not possible." During the final steps of pump assembly, Entergypersonnel used a number of cloth FME covers to prevent nuts and washers from fallinginto the open piping. Because the area was not designated a FME Zone 1, the clothcovers were not tied down or logged as FME zone inventory, and one cover was leftbehind in the system after the pump was completely installed. During post-maintenancetesting on December 30, 2010, Entergy observed that the pump did not meet the flowrate acceptance criterion that is required for operability. On January 2,2011, the newlyinstalled pump was removed for internal inspection, and a cloth FME cover was foundlodged in the pump. Part of the cover had been torn away during the pump run andcariied further into the RHRSW system. Subsequent system inspection identified alarge piece of the cover on the 'A' Residual Heat Removal Heat Exchanger (RHR HX)baffle plate and small pieces in other areas of the 'A' RHRSW train. Discovery of thismaterial in the'A'RHR HX rendered the entire RHRSW'A'train inoperable as ofDecember 30, when the unacceptable flow rate was first discovered. Entergysubsequently removed all of the foreign materialfrom the 'A' RHRSW train. On January7, 2011, Entergy successfully tested the 'A' RHRSW train and returned it to service. The'C' RHRSW pump was successfully tested and returned to service on January 8, 2011.This issue was entered into Vermont Yankee's corrective action program. Shortly afterretrieval of the FME cover, Entergy conducted a "stand down" to discuss the event andreinforce FME control standards. lmmediate corrective actions included conducting a"stand down," reinforcing the standards and requirements for FME controls and generalprocedural compliance, as well as reinforcing expectations for the attention to detail ofwork practices. Additionally, Entergy entered the deficiency into their corrective actionprogram to evaluate for additional corrective measures.Enclosure 13Analvsis: The performance deficiency was that Entergy did not fully implement writtenprocedures, as required by Technical Specification 6.4 and Entergy procedure EN-MA-118, covering preventive and corrective maintenance operations which could have aneffect on the safety of the reactor. Specifically, Entergy performed the closeoutinspection prior to RHRSW system closure, and did not establish a FME Zone l duringthe remaining work activities prior to system closure. This issue was within Entergy'sability to foresee and correct and should have been prevented. This led to foreignmaterial intrusion into the'A'train of RHRSW, rendering the'A'train inoperable.Traditional Enforcement did not apply; as the issue did not have actual or potentialsafety consequences, had no willful aspects, nor did it impact the NRC's ability toperform its regulatory function. A review of NRC IMC 0612, Appendix E, "MinorExamples," revealed that no minor examples were applicable to this finding. Theinspectors determined that the finding was more than minor because it was associatedwith the Equipment Performance attribute of the Mitigating Systems cornerstone, andaffected the cornerstone objective to ensure the availability of systems that respond toinitiating events to prevent undesirable consequences, (i.e., core damage). Specifically,materialfrom the FME cover made its way into the 'A' RHR HX and rendered the'A'RHRSW train inoperable for greater than 7 days. A review of NRC IMC 0612, AppendixE, "Minor Examples," revealed that no minor examples were applicable to this finding.The inspectors used IMC 0609.04, "Phase 1 - InitialScreening and Characterization ofFindings," and determined that the finding required a Phase 2 review because the'A'RHRSW train had an actual loss of safety function for greater than its allowed outagetime (7 days). Using IMC 0609 Appendix A, "Determining the Significance of ReactorInspection Findings for At-Power Situations," and an event likelihood of 3-30 days, theinspectors determined that the finding was of very low safety significance (Green). Themost dominant core damage sequence was a transient without the power conversionsystem (TPCS): TPCS(1) + cHR(2) + CV(3) = 6 (Green). The risk was mitigated bythe unaffected 'B' RHR heat exchanger and by the containment vent'The finding had a cross-cutting aspect in the Human Performance cross-cutting area,Work Practices component, because Entergy personnel did not follow procedure EN-MA-118. Specifically, Entergy failed to establish a FME Zone 1 after the systemcloseout inspection. H.4(b)]Enforcemerf[ Technical Specification 6.4, "Procedures," requires that writtenprocedures be implemented for activities including "preventive and correctivemaintenance operations which could have an effect on the safety of the reactor."Contrary to the above, the requirements of EN-MA-118, "Foreign Material" were not fullyimplemented during the pump assembly portion of the work activity. This led to foreignmaterial intrusion into the 'A' RHRSW train that rendered it inoperable from December30, 2010 to January 7, 2011 . lmmediate corrective actions included conducting a "standdown," reinforcing the standards and requirements for FME controls and generalprocedural compliance, as well as reinforcing expectations for the attention to detail ofwork practices. Additionally, Entergy entered the issue into their corrective actionprogram to evaluate for additional corrective measures. Because this finding is of veryEnclosure 14low safety significance and Entergy has entered it into their corrective action program(CR-WY-2011-0007), this violation is being treated as an NCV, consistent with the NRCEnforcement Policy. (NCV 0500027112011002-01: Failure to Follow Foreign MaterialExclusion Procedure)1R22 Surveillance Testino (71111.22)a. lnspection Scope (6 samples)The inspectors observed six surveillance tests and/or reviewed test data of selected risk-significant SSCs to determine whether the testing adequately demonstrated equipmentoperational readiness and the ability to perform the intended safety functions. Theinspectors reviewed selected prerequisites and precautions to determine if they weremet; evaluated whether the tests were performed in accordance with the writtenprocedure; determined whether the test data was complete and met proceduralrequirements; and assessed whether SSCs were properly returned to service followingtesting. The inspectors also verified that conditions adverse to quality were entered intothe CAP for resolution. The documents reviewed are listed in the Attachment. Theinspectors reviewed the following surveillance tests:. 'A' Emergency Diesel Generator Monthly Surveillance;. Service Water Pump Testing;. 'B' Loop RHFJRHRSW Pump and Valve Operability and Full Flow Test;. Main and Auxiliary Steam System Surveillance;r Quarterly Main Turbine Valve Performance Testing; ando Reactor Coolant System Leak Detection Surveillance (RCS LD).b. FindinqsNo findings were identified.Cornerstone: Emergency Preparedness (EP)1EP6 Drill Evaluation (71114.06)Emeroencv Preparedness Drilla. Inspection Scope (2 samples)The inspectors observed an emergency preparedness (EP) drill on January 19,2411,and observed the player critiques. Entergy's EP staff preselected the drill notificationsand protective action recommendations to be included in the EP drill performanceindicator (Pl). The inspectors discussed the performance expectations and results withEntergy's EP staff to confirm correct implementation of the Pl program. The inspectorsfocused on the ability of licensed operators to perform event classifications and theability of designated personnel to make proper notifications in accordance with Entergy'sEnclosure 15procedures and industry guidance. The inspectors evaluated the drillfor conformancewith the requirements of 10 CFR Part 50, Appendix E, "Emergency Planning andPreparedness for Production and Utilization Facilities." The inspectors comparedEntergy's self-identified issues with observations from the inspectors' review to ensurethat performance issues were properly identified and documented. The documentsreviewed are listed in the Attachment.The inspectors observed licensed operator "as found" simulator training on February 7,2011. The inspectors evaluated the operating crew activities related to accurate andtimely classification and notification of an Alert. Additionally, the inspectors assessedthe critique process used by the training evaluators for its ability to identify performancedeficiencies. The documents reviewed are listed in the Attachment.These activities constituted two drill evaluation inspection samples.b. FindinqsNo findings were identified.4. OTHER ACTTVTTES IOAI4OA1 Performance Indicator (Pl) Verification (71151- 3 samples)lnitiatino Events CornerstoneInspection ScopeThe inspectors reviewed Entergy's submittals and Pl data for the cornerstones listedbelow for the period from January 201A to December 2010. The inspectors reviewedselected operator logs, plant process computer data, licensee event reports, andcondition reports. The Pl definitions and guidance contained in Nuclear Energy Institute(NEl) 99-02, "Regulatory Assessment Performance lndicator Guideline," Revision 6, EN-Ll-1l4, "Performance Indicator Process," Revision 4, and AP 0094, 'NRC Performancelndicator Reporting," Revision 15, were used to verify the accuracy and completeness ofthe Pl data reported during this period. The Pls reviewed were:. Unplanned scrams per 7000 critical hours;. Unplanned power changes per 7000 critical hours; andr Unplanned scrams with complications.FindinqsNo findings were identified.a.b.Enclosure 164OA2 ldentification and Resolution of Problems (71152).1 Reviews of ltems Entered into the Corrective Action Proqrama. Inspection ScopeThe inspectors performed a daily screening of each item entered into Entergy's CAP.This review was accomplished by reviewing printouts of each CR, attending dailyscreening meetings, and/or accessing Entergy's database. The purpose of this reviewwas to identify conditions such as repetitive equipment failures or human performanceissues that might warrant additional follow up.b. FindinqsNo findings or observations were identified..2 Operator Workaroundsa. Inspection Scope (1 sample)The inspectors reviewed the cumulative effect of operator workarounds, operatorburdens, enhanced surveillances and control room deficiencies on the reliability,availability and potential mis-operation of mitigating systems with a particular focus onissues that had the potential to affect the ability of operators to respond to planttransients and events. The inspectors reviewed the auxiliary operator roundsheets/turnover sheets for the reactor building, turbine building, and outside areas of theplant, and compared these with Entergy's listed operator burdens and workarounds.The inspectors reviewed selected off-normal procedures and walked down related areasof the plant to determine whether the procedure steps could be implemented byoperations personnel and required equipment was properly staged. ln addition, theinspectors reviewed Entergy tracking systems for operator burdens, control roomdeficiencies, and disabled control room alarms. The inspectors discussed selectedissues with responsible operations personnel to ensure they were appropriatelycategorized and tracked for resolution.b. FindinqsNo findings or observations were identified.4OA3 Event Follow-up (71153)Plant Event ReviewInspection Scope (1 sample)On February 16, 2011, while performing the quarterly surveillance test on the HighPressure Coolant system (HPCI) turbine, a steam leak developed at the flange on steam.1Enclosure b.17trap 23T-3 after full steam line pressure was applied to the trap during the test. HPCIroom temperatures increased causing localfire alarms to activate. Based on the rapidrise in temperature in the HPCI room, operators manually isolated the HPCI system.This action occurred before the room temperatures reached the automatic isolation setpoint for the HPCI system. The inspectors observed plant parameters from the controlroom and reviewed control room operator performance. The inspectors communicatedthe plant event to regional personnel and compared the event details with criteriacontained in IMC 0309, "Reactive Inspection Decision Basis for Reactors," forconsideration of additional reactive inspection activities. The inspectors reviewedEntergy's corrective actions to ensure they were implemented commensurate with theirsafety significance.Findinqs and ObservationsIntroduction: A self-revealing, Green NCV of Technical Specification 6.4, "Procedures,"was identified in which maintenance and planning personnel did not involve engineeringpersonnel as required by EN-MA-101, "Fundamentals of Maintenance," Revision 9, andEN-WM-105, "Planning," Revision 8, resulting in the incorrect material being used toreplace the gasket on the flange of HPCI steam trap 23T-3. Entergy ultimately replacedthe gasket with the correct material and entered this issue into their corrective actionprogram.Description: On February 1, 2011, the HPCI system was removed from service to repaira small steam leak in non-safety related one-inch piping downstream of steam trap 23T-3. The flange on the trap had to be disassembled to access and replace the piping withthe steam leak. The flange was originally sealed with a spiral wound flexitallic gasket.This type of gasket was not readily available and the licensee determined that a Garlock9920 gasket was an acceptable replacement. The decision was made by maintenancesupervision based on a previous Technical Evaluation (04-00600 revision 0) provided inthe work package by the planning department. This technical evaluation states that thismaterial should not be used in systems greater than 250 psig. This limitation wasoverlooked and the Garlock 9920 gasket was put into place on 23T-3. Entergyprocedure EN-MA-101 states that replacement components shall be "like for like," andEN-WM-105 states that the Procurement Engineering Group (PEG) be notified if itemscannot be verified by procedure or EN-DC-313, "Procurement Engineering Process,"Revision 5. Neither procedure was followed for the replacement gasket in this instance.After replacing the steam trap flange gasket with Garlock 9920, the HPCI system wasrestored to standby status. Work Order (WO) 252692 required the piping and flange betested for leakage at full system pressure (approximately 1000 psig). The postmaintenance test (PMT) listed in the work order did not provide the operationsdepartment with detailed guidance in establishing initial conditions for the test.Operators believed that the steam trap gasket was at the required PMT pressure whenaligned to the standby configuration. However, with HPCI in a standby configuration, aseries of two normally-opened isolation valves provided a drain pathway to the maincondenser hotwell environment. Due to the low pressure condition at the steam trapEnclosure 18flange gasket, the PMT had been inappropriately considered satisfactory, and Entergydeclared the HPCI system to be operable on February 1.On February 16, during HPCI quarterly surveillance testing, the steam trap andassociated piping were exposed to full HPCI system steam pressure because theisolation valves to the main condenser automatically closed as part of the HPCI start-upsequence for the post-maintenance testing. The new gasket failed when exposed topressure beyond its design rating, and allowed steam to escape between the flange andthe steam trap body. The amount of steam that issued from 23T-3 was substantialenough to fill the room and raise the ambient temperature. Auxiliary operators in theHPCI room immediately reported the steam leak to the main control room, wherelicensed operators remotely isolated the HPCI steam line to stop the flow of steam.This deficiency was entered into Entergy's corrective action program as CR-WY- 2011-00667. Entergy determined that the root cause of the event was determined to be theincorrect use of the Garlock 9920 materialfor the gasket. Additionally, Entergydetermined that inadequate post maintenance testing was a contributing cause. OnFebruary 18,2011, Entergy replaced the 23T-3 flange gasket with the appropriatematerial, and completed a successful post maintenance test.Analvsis: The inspectors determined that the installation of inappropriate material for thesteam trap flange gasket was a performance deficiency which caused the HPCI systemto be inoperable for greater than the time allowed by Technical Specifications. Thisperformance deficiency was within Entergy's ability to foresee and correct and shouldhave been prevented. Traditional enforcement does not apply as the issue did not havean actual safety consequence, had no willful aspects, nor did it impact the NRC's abilityto perform its regulatory function.The inspectors reviewed Inspection IMC 0612, Appendix E, "Minor Examples," anddetermined that this deficiency was not similar to any of the minor examples.Additionally, using IMC 0612, "Power Reactor Inspection Reports," Appendix B, theinspectors determined that the finding was more than minor because it was associatedwith the Human Performance attribute of the Mitigating Systems cornerstone, andaffected the cornerstone objective to ensure the availability of systems that respond toinitiating events to prevent undesirable consequences. The finding was determined tobe of very low safety significance in accordance with IMC 0609, Appendix A,"Determining the Significance of Reactor Inspection Findings for At-Power Situations,"using significance determination process (SDP) Phases 1,2 and 3. Phase 1 screenedthe finding to Phase 2 because it represented an actual loss of the HPCI system safetyfunction. A Region I Senior Reactor Analyst (SRA) conducted a Phase 3 analysisbecause the Phase 2 analysis, conducted by the inspectors using the W Pre-solvedRisk-lnformed Inspection Notebook, indicated that the finding had the potential to begreater than very low safety significance (Greater than Green).The SRA used the W Standardized Plant Analysis Risk (SPAR) model, Revision 8.16,to conduct the Phase 3 SDP evaluation, assuming that HPCI would not have been ableto perform its safety function over the 19 day period from February 1 , 2011 to FebruaryEnclosure

.219 19,2011. This analysis indicated an increase in core damage frequency (ACDF) forinternal initiating events in the range of 1 core damage accident in 4,000,000 years ofreactor operation; in the low 1E-7 range per year. The dominate core damagesequences included the operator failure of HPCI and reactor core isolation cooling(RCIC), and the failure of operators to depressurize the reactor following a loss of mainfeedwater. ln accordance with IMC 0609, for a finding with an internal events ACDFgreater than 1E-7, the SRA assessed the impact of the finding on: 1) External eventssuch as fire, seismic and flooding, determining, based on review of the W IndividualPlant Examination for External Events, that the total ACDF (internal plus external) wouldnot be above the 1 E-6 threshold; and 2) the increase in large early release frequency(ALERF), determining that given the operators ability, following core damage, todepressurize and inject water to the reactor from low pressure sources and to flood thecontainment that the ALERF was in the low E-8 range. The Phase 3 SDP analysisdetermined that this issue was of very low safety significance (Green).This issue has been entered into Vermont Yankee's corrective action program. Theflange gasket for 23T-3 was immediately replaced with the correct material. Personnelinvolved in the event were coached on procedures for substituting material andcomponents.This finding had a cross-cutting aspect in the Human Performance cross-cutting area,Decision Making component, because Vermont Yankee personnel did not obtaininterdisciplinary input on the decision to use a different, incorrect gasket material in asteam trap in the HPCI system. H.1(a)]Enforcement: Technical Specification 6.4, "Procedures," requires that writtenprocedures be implemented for preventive and corrective maintenance operations thatcould have an effect on the safety of the reactor. Contrary to this requirement, onFebruary 1, 2011, the requirements of EN-MA-101, "Fundamentals of Maintenance," aswell as, EN-WM-105, "Planning," were not properly implemented. Specifically, Entergyperformed corrective maintenance to replace a HPCI system gasket that was not "like forlike" (contrary to EN-MA-101), and the Procurement Engineering Group was not notifiedfor the use of a new type of item (contrary to EN-WM-105). This action led to the HPCIsystem being inoperable from February 1,201 1 to February 19, 2011. lmmediatecorrective actions included installation of the proper gasket, followed by successfulcompletion of a proper post-installation pressure test of the gasket. Because of the verylow safety significance (Green) and because it has been entered into the CAP (CR-VTY-2011-00667), the NRC is treating this finding as a NCV, consistent with the NRCEnforcement Policy. (NCV 0500027112011002-02: Steam Leak on High PressureCoolant Injection (HPCI) During Surveillance Testing)(Closed) LER 05000271/2010-002-00&01: Inoperabilitv of Main Steam Safetv ReliefValves Due to Deqraded Thread Seals (71153 - 1 sample)During the 2010 refueling outage, the pneumatic actuators for the four main steamsafety relief valves (SRVs) were tested and leakage was identified through the shaft-to-piston thread seal that was in excess of the design requirement on two of the four SRVs.Enclosure

20Material testing determined that the apparent cause of the degraded thread sealcondition was thermal degradation. The thread seals were replaced and tested on allfour SRVs prior to startup from the 2010 refueling outage.Entergy determined that this potentially affected the ability of the SRVs to perform theirmanual and automatic depressurization function, as required by TechnicalSpecifications, since the leakage impacted the ability of the SRVs to satisfy designactuation requirements. Entergy determined that there was firm evidence that thiscondition may have existed for a period of time greater than allowed by TechnicalSpecifications, and therefore this event was reportable.Due to the availability of a safety-class back-up nitrogen supply with separate pressureregulators, Entergy determined that adequate capacity for the AutomaticDepressurization System (ADS)existed at all times. Due to the redundancy in ADSdesign, the availability of the HPCI system, and the availability of a safety-class backupnitrogen supply, the ability to depressurize the reactor was maintained, and there was nopotential adverse impact to public health and safety.The inspectors reviewed the subject LER, the as-found condition during the refuelingoutage, the subsequent material testing and analysis, and Entergy's evaluation of thecondition. A violation of very low safety significance (Green) was identified by thelicensee. The enforcement aspects of this finding are discussed in Section 4OA7. ThisLER is closed.4OAO Meetinqs, includinq ExitExit Meetino SummarvOn April 11, 2011 , the resident inspectors presented the first quarter inspection results toMr. Michael Colomb, Site Vice President, and other members of the Vermont Yankeestaff. The inspectors confirmed that any proprietary information provided or examinedduring the inspection had been returned to the licensee.4C.A7 Licensee-ldentified ViolationsThe following violations of very low safety significance (Green) were identified by thelicensee and are violations of NRC requirements, which meet the criteria of the NRCEnforcement Policy for being dispositioned as non-cited violations..1 Technical Specification 3.5.F, "Automatic Depressurization System," allows up to one offour SRVs in the automatic depressurization system to be inoperable for up to sevendays at any time the reactor steam pressure is above 150 psig with irradiated fuel withinthe vessel, or an orderly shutdown of the reactor shall be initiated and the reactorpressure shall be reduced to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Contrary to the above,Entergy determined that two (2) of the four (4) SRVs were inoperable for a period of timegreater than allowed by Technical Specifications. This determination was based onpneumatic actuator thread seal leakage that was identified during testing of theEnclosure

,221pneumatic SRV actuators in the 2010 refueling outage. Entergy determined the leakageto be in excess of design requirements. This condition has been entered in thelicensee's corrective action program (CR-WY-2O10-2187) and corrective actions havebeen developed.The inspectors determined that this finding was more than minor because it adverselyaffected the Mitigation Systems cornerstone objective of ensuring the reliability ofsystems that respond to initiating events to prevent undesirable consequences. Theinspectors determined that the function for core decay removalwas affected, since thesafety function of the ADS valves is to depressurize the reactor to allow for low pressurecoolant injection. The inspectors determined that this finding was not greater thanGreen, because subsequent laboratory analysis and engineering evaluation documentedin Entergy Operability Recommendation WY 2011-0631 concluded that sufficientmargin was available in the safety-class backup supply to the pneumatic actuationsystem. The inspectors reviewed Entergy's laboratory results and OperabilityRecommendation, and concluded that the ADS function would have been met under theworst case leakage for all design basis conditions.Technical Specification 3.6.D, "Safety and Relief Valves," requires the reactor to be shutdown and pressure brought below 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with two (2) or more SRVsinoperable. Contrary to the above, Entergy determined that two (2) of the four (a) SRVswere inoperable for a period of time greater than allowed by Technical Specifications.This determination was based on pneumatic actuator thread seal leakage that wasidentified during testing of the pneumatic SRV actuators in the 2010 refueling outage.Entergy determined the leakage was in excess of design requirements, therebyrendering the SRV manual depressurization function inoperable. This condition hasbeen entered in the licensee's corrective action program (CR-WY-2010-2187) andcorrective actions have been developed.The inspectors determined that this finding was more than minor because it adverselyaffected the Mitigation Systems cornerstone objective of ensuring the reliability ofsystems that respond to initiating events to prevent undesirable consequences. Theinspectors determined that the function for core decay heat removal was atfected, sincethe ability to manually discharge steam from core decay heat to the suppression poolwas degraded by the thread seal leakage. The inspectors determined that this finding isnot greater than Green, because subsequent laboratory analysis and engineeringevaluation documented in Entergy Operability Recommendation VTY 2011-0631concluded that sufficient margin was available in the safety-class backup supply to thepneumatic actuation system. The inspectors reviewed Entergy's laboratory results andOperability Recommendation, and concluded that the SRV manual depressurizationfunction would have been met under the worst case leakage for all design basisconditions.10 CFR 50.65(a)(4) requires, in part, that before performing maintenance activities, thelicensee shall assess and manage the increase in risk that may result from proposedmaintenance activities, Contrary to the above, on January 3,2011, Entergy did notadequately assess and manage the increase in risk due to proposed emergent.3Enclosure 22maintenance activities. This resulted in a non-conservative risk assessment and failureto take all of the appropriate risk management actions for the actual plant conditions.Entergy identified this after the emergent maintenance activities had been completed,and entered the issue into their corrective action program (CR-WY-2011-00028) toevaluate for appropriate corrective actions. The finding is more than minor because it issimilar to IMC 0612, Appendix E, Example 7.e; in that, the overall elevated plant risk putthe plant in a higher licensee-established risk category. The finding was evaluated usingIMC 0609 Appendix K, "Maintenance Risk Assessment and Risk ManagementSignificance Determination Process," and was determined to be of very low safetysignificance (Green) because the Incremental Core Damage Probability Deficit betweenthe actual plant conditions and the incorrect risk assessment for the duration of theactivity was less than 1.0 E-6 (approximately 3.3 E-9).ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Vermont Yankee PersonnelM, Colomb, Site Vice President

C. Wamser, General Manager of Plant Operations

M. Romeo, Director of Nuclear Safety

R. Wanczyk, Licensing Manager

N. Rademacher, Director of Engineering

M. Gosekamp, Operations Manager

J. Rogers, Design Engineering Manager

J. Merkle, System Engineering Manager

D. Jones, Asst. Operations Manager

P. Ryan, Security Manager

B. Pittman, Assistant Operations Manager

M. Tessier, Maintenance Manager

J. Hardy, Chemistry Manager

P. Corbett, Quality Assurance Manager

S. Naeck, Outage Manager

J. Bengtson, CA&A Manager

M. Castronova, Manager of Projects

J. Ward, l&C Superintendent

R. Heathwaite, Chemistry Supervisor

C. Daniels, FIN Team Superintendent

R. Current, Sr. Electrical l&C System Engineer

L. Doucette, System Engineer

J. Devincentis, Licensing Engineer

P. Couture, Licensing Specialist

J. Meyer, Licensing Specialist

M. Morgan, Technical Training Superintendent

M. Anderson, Fire Protection Engineer

M. Pletcher, Shift Technical Advisor

K. Oliver, Shift Manager

V. Ferrizzi, Shift Manager

J. Miller, Auxiliary Operator

J. Kritzer, Shift Technical Advisor

D. Hensel, Work Week Manager

F. Aldrich, Control Room Supervisor

N. Jennison, Shift Manager

G. Bacala, Control Room Supervisor

J. Clough, System Engineer

D. Macie, Facilities

S. Nelson, Fire Brigade lnstructor

J. Stasolla, Mechanical Systems Engineer

B. Pelzer, Code Programs EngineerAttachment

A. Robertshaw, Mechanical Design Engineer

P. Jerz, Work Week Manager

J. Devine, Auxiliary Operator

S. Jonasch, Mechanical Systems EngineerOpened and Closed05000271/201 1002-0105000271/2011002-02Closed0500027 1 120 1 0-002-00&01NCVNCVA-2Failure to Follow Foreign Material Exclusion Procedure(Section 1R19)Steam Leak on High Pressure Coolant lnjection (HPCI)During Surveillance Testing (Section 4OA3)

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

LERInoperability of Main Steam Safety Relief ValvesDue to Degraded Thread Seals (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

ln addition to the documents identified in the body of this report, the inspectors reviewed thefollowing documents and records:Vermont Yankee Nuclear Power Station Updated Final Safety Analysis ReportVermont Yankee Nuclear Power Station Technical SpecificationsVermont Yankee Nuclear Power Station Narrative Logs, Night Orders, and Standing Orders

Section 1R01: Adverse Weather ProtectionProceduresOP 3127, "Natural Phenomena," Rev. 26Condition ReportsCR- 2011-00946, "Drain in Stairwell of Admin Building North Exit is Plugged"CR-2011-00948, "Water is Leaking from Cracks in the Concrete Ceiling"CR-2005-02008. "Water Found in East

SWGR Room"Attachment

A-3

Section 1R04: Equipment AlisnmentProceduresOPST-CS-4123-OOa, "Core Spray Pump'A' Comprehensive Operability Test," Rev. 2OP 2122, "Auto Blowdown System," Rev. 23OP 2123, "Core Spray," Rev. 43OP 2126, "Diesel Generators," Rev. 26OP 3122, "Loss of Normal Power, "Rev. 42OP 3126, "Shutdown Using Alternate Shutdown Methods," Rev. 42OP 4107, "EoP/Alternate Shutdown Tools and Supplies Surveillance," Rev. 14OP 2124, "Residual Heat Removal System," Rev. 114OPOP-4kv-zl42, "4kv Electrical System," Rev. 00Condition ReportsCR -2010-1440, "STA Switch Assembly Screw Backing Out"CR-2011-801, "FCV-6-128 is not Open as Much as Expected"Drawinos5920-04150, "Schematic Lube Oil System," Rev. 95920-04147, "Pl&D Emergency Diesel Generator

DG-1-1B Air Jacket Coolant System," Rev. 0,Sheet 2G-191160, "Flow Diagram Diesel Generator Standing Air System," Rev. 23, Sheet 7G-191167, "Flow Diagram Nuclear Boiler," Rev. 76G-191299, "4kv Auxiliary One Line diagram," Rev. 31G-191 159, "Flow Diagram Service Water System," Rev. 81, Sheet 1Miscellaneous Documents"Design Basis Document for Safety Related 4.16 kVl480V System," Rev. 254k Volt AC System Health Report - 3'o Quarter 2010EMST-RLAY-4256-01, "Calibration of Degraded Grid Area RXKEI Timing Relays Switchgear3," Rev. 00Work Orders00231773, "'C' RHRSW Pump Control Switch Problems"

Section 1R05: Fire ProtectionProceduresEN-OC-127, "Control of Hot Work and lgnition Sources," Rev. 8AP 0042, "Plant Fire Prevention and Fire Protection," Rev. 53AP Q077, "Barrier Control Process," Rev. 20DrawinqsG-191163, "Flow Diagram Fire Protection System lnner Loop," Rev. 44, Sheet 1Miscellaneous DocumentsFire Hazards Analysis App. B, Rev. 11VY

SSCA "Safe Shutdown Capability Analysis" Vol. 1, Rev. 9PFP-T-TB-8 "Fire Brigade Pre-Fire Plans - Lube Oil Room," Rev. 0PFP-CB-2 "Fire Brigade Pre-Fire Plans - Cable Vault," Rev. 0PFP-TRAN "Fire Brigade Pre-Fire Plans - Transformers," Rev. 0Attachment

SIP-11-02 "Fire Protection System lmpairment Permit,BCP-201 1-07, "Barrier Control Permit for HPCI Door"SIP-11-02 "Fire Protection System lmpairment Permit," 1110111BCP-201 1-07, "Barrier Control Permit for HPCI Door"BCP-2011-13 "Doors propped open for major diesel overhaul"SIP-2010-55, "Fire Protection System lmpairment Permit for Southwest Corner Room"FCBT-SAF-Firewatch, "Hot Work Firewatch Training," Rev. 0Work Orders52258739, "OP 4019 (SA) Perform Door Inspection"Condition ReportsCR-WY-20 1

1-00872, "Fire Watch Tour Expectations"

Section 1R06: Flood Protection MeasuresCalculationsWC-1774, "Flooding from 4" Fire Protection Pipe Break in the Reactor Building Elevation 252feet 6 inches," Rev. 0VYC-1787,"Flooding from Service Water Pipe Break in the Reactor Building," Rev. 1MiscellaneousVermont Yankee Internal Flooding Topical Design Basis Document, Rev. 9VY-NE-09-0001, "lnternal Flooding Analysis," Rev. 0Section 1Rl 1: Licensed Operator Requalification ProgramProceduresOP 3511, "Off-Site Protective Action Recommendations," Rev.27OP 3540, "Control Room Actions during an Emergency," Rev. 25MiscellaneousNEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Rev. 5AFG 42,'As-found Simulator Evaluation Guide," Rev.1Section 1R12: Maintenance EffectivenessCondition ReportsCR-WY-2009-03101 , "Degraded Wall Thickness on Instrument Air Dryer

D-1-1 B Piping"CR-WY-2009-03144, "Wall Thinning ldentified on D-1-1 B"CR-\flY-2010-05214, 'OG-208A, Offgas Inlet Valve Will Not lsolate"CR-WY-2O1 0-03884, "Untimely Repair of

AOV-OG-1 01A"CR-WY-2008-02006, "AOG - Equipment Train 'A' Now above Maintenance Rule ReliabilityCriteria"CR-WY-2010-03971, "Due Date Extension Approved Without Director Review"CR-WY-2O10-01906, "'C' Station Air Compressor Tripped"CR-WY-2O10-05425,"C-1-1B Service Air Compressor Tripped on Low Oil Pressure"CR-WY-2O11-00477, "'B' Service Air Compressor Tripped on Oil Pressure"Work OrdersWO

00235150, "C-1-1C, Troubleshoot BreakerTrip per

EN-MA-125"ProceduresEN-DC-206, "Maintenance Rule (aX1) Process," Rev. 1Attachment

A-5EN-DC-207, "Maintenance Rule Periodic Assessment," Rev. 2EN-DC-204, "Maintenance Rule Scope and Basis," Rev. 2EN-DC-205, "Maintenance Rule Monitoring"Miscellaneous DocumentsVYSE-MRL-2008-013, "Performance Evaluation for AOG Equipment Train A," Rev. 1AOG, "Augmented Offgas Maintenance Rule SSC Basis Document," Rev. 3AOG Preventive Maintenance Task ListAOG SSC Performance History

11112008 - 111412011AOV-OG-101A Action Plan, updated 10/06/10Maintenance Rule Monthly Report for December 2010EN-LI-102, "Corrective Actions Process," Rev. 16lA, "lnstrument Air Maintenance Rule Scoping Basis Document," Rev. 3State of the System Report - Instrument Air, 113112011VYSE-MRL-2010-030, "Performance Evaluation for lnstrument Air System Train 'A'," Rev. 0SA, "Seryice Air Maintenance Rule Scoping Basis Document," Rev. 5State of the System Report - Service Air, 113112011DrawinqsDWG-3360 0-A-207, "Engineering Flow Diagram Train'A' Recombiner Area OffgasModification," Rev. 27

Section 1Rl3: Maintenance Risk Assessments and Emergent Work GontrolProceduresAP Q172, "Work Schedule Risk Management - Online," Rev. 22EN-OP-119, "Protected Equipment Postings," Rev. 1EN-OP-1 19, "Protected Equipment'Postings," Rev. 2OP-4114, "Standby Liquid Control Surveillance," Rev. 66Condition ReportsCR-WY-2O11-00388 Additional Plant Equipment Requiring Protection per

EN-CP-119ldentified LateCR-WY-2011-00445 QTR

OP-4181 Service Water Valve Operability Testing Was DelayedDuring WW1105CR-WY-2O11-00028 Unanticipated Change in EOOS Risk Color during WW1101CR-WY-2O11-01184, "standby Liquid Control not Declared Unavailable during Surveillance"Miscellaneous Documents"VY EOOS Risk Assessment - WW1103,"Rev. 2EOOS Risk Assessment ToolEMMP-INSP-00216-22, "Weekly Yard Reading and Brush Inspection," Rev. 3Online Maintenance Safety Assessment Review

1131111 - 217111Online Maintenance Safety Assessment Review 212111VYAPF 017 2.O1 "Online Maintenance Safety Assessment Review 1 l2l 1 1 -1 l 41 1 1Online Maintenance Safety Assessment Review 2116111Attachment

A-6NUMARC 93-01 Section 1 1, "Assessment of Risk Resulting from Performance of MaintenanceActivities," Rev.2Work Week 1111 Schedule

Section 1R15: Operabilitv EvaluationsProceduresEN-OP-104, "Operability Determination Process," Rev. 4CHOP-DIES-4613-01, "Sampling and Testing of Diesel Fuel Oil," Rev. 0EN-OP-104, "Operability Determination Process," Rev. 5VYEM 107, "Emergency Diesel Generators Service Manual," Rev. 17Miscellaneous DocumentsASTM

D-975-00, "Standard Specification for Diesel Fuel Oils"SC 11-01, "GE Hitachi 10CFR Part21 Communication," dated February 15,2011ODMI, "Crack Indications in Marathon Control Blades have Been Observed in an lnternationalBWR," Rev. 1Section 1R{8: Plant ModificationsDrawinqsG

191173, "Flow Diagram Fuel Pool Cooling and Clean Up System," Sheet 2, Rev. 9Miscellaneous DocumentsEC 21288, "Replace V76-38 with New Check Valve"EC lT444, "ChemicalTreatment Connections to the Spent Fuel Pool Cooling System"Work OrdersWO

52207081, "service Water Check Valve Inspection for Swing Type Check Valve"WO

00229506, "Replace Check Valve V76-38"Section I Rl 9: Post-Maintenance TestinsProceduresOP 4124, "Residual Heat Removal and RHR Service Water System Surveillance," Rev. 117OP 4124, "Residual Heat Removal and RHR Service Water System Surveillance," Rev. 119EN-WM-107, "Post Maintenance Testing," Rev. 2EN-MA-118 "Foreign Material Exclusion," Rev. 7OP 4181, "Service Water," Rev. 73VYOPF 4181.08, "Service Water Pump Capacity Test" completed 03/10/11 and 03112111VYOPF 4181.04, "Service Water Pump Capacity Test Data Sheet" completed 03112111VYOPF 4124.04A, 'RHR Pump 'A' (P-10-1A) Operability Data Sheet" completed 01107111VYOPF 4124.04C, "RHR Pump'C' (P-10-1C) Operability Data Sheet" completed O1lO7l11VYOPF 4124.06A, "RHRSW Pump 'A'(P-8-1A) and Valve Operability and Full Flow Test DataSheet" completed 01 lO7 /1 1OP 4126, "Diesel Generators Surveillance," Rev. 85Op 2180, "Circulating Water/Cooling Tower Operation," Rev. 99ECT 15732-01, Rev. 00Attachment

A-7Condition ReportsCR-WY-2011-00067, "Unexpected Annunciator "RHR Pump A Seal LKG Hl" Locked ln"CR-WY-2011-00007, 'P-8-1C: Discovered 24" FME Cover Lodged in Lower lmpeller duringPump Removal"CR-WY-2O 1 1 -01 054, "lncorrect Valve Positioned during Surveillance"CR-WY-2011-01325, "Total Dynamic Head Anomaly when Testing RHRSW Pump P-8-18"Work OrdersWO

52207081, "Service Water Check Valve Inspection for Swing Type Check Valve"WO

00229506, "Replace Check Valve V76-3B'WO

52294475, "Replace'B'Service Water Pump"WO

52290650, "Drain Hydro Diesel Generator Jacket Cooling System"WO

52290096, "Major Diesel Overhaul and Inspection"WA

00244609, "DG-1-1B: Replace Blower Cover Gasket"WO

52290648, "Diesel Generator Temperature Control Valve Refurbishment"WO

52290258, "DG-1-18: Replace or Rebuild the M2 and M5 Contactors"WO

00253892, "DG-B: Replace Aftercooler HX Floating Channel Head"WO

00258475, "Small Air Leak on 'B' EDG Starting Air Compressor"WO

00252692, "ST-23-3, Replace /Repair Steam Trap"WO

52212754,"Circ Water Pump Overhaul"WO

00200034, "Replace'B' RHRSW Pump with New Pump from Hayward Tyler"Section 1 R22: Surveillance TestinqCondition ReportsCR-VTY-2010-00469, 'RHR System Unavailability during RHR Valve Surveillance"CR-\ffY-2010-04810, "DG-1-1A Lube Oil Leak of Approx. 30 DPM Observed"CR-WY-2010-05129, "Approx. 20 DPM Lube Oil Leak on'A'EDG Lube Oil HX"CR-WY-2011-00097, "Minor Lube Oil Leak on 'A' EDG Lube Oil HX South End"ProceduresOP 4113, "Main and Auxiliary Steam System Surveillance," Rev. 34OP 4124, "Residual Heat Removal and RHR Service Water System Surveillance," Rev, 117OP 4181, "service Water/Alternate Cooling System Surveillance," Rev. 73VYOPF 4184.01, "station Service Water Pump Operability Test," completed 02103111VYOPF 4126.Q2, "Diesel Generator Operating Data," completed

12113110 and 01l10111VYOPF 4126.13, "Diesel Generator Slow Start Operability Test," completed

12113110 and01t10111Miscellaneous DocumentsEDG "Emergency Diesel Generators and Auxiliary Systems Design Basis Document," Rev.22IST Component Basis - Pumps, Rev. 13ML031780796 Safety Evaluation for Relief Requests Related to the Fourth 1O-Year ISTProgram Service Water Pump P-7-1A, B, C, D Test Curve, 4113110ESOM Operator Rounds Logs from November 26,2010 - February 23,2011 Stations 15 and 16Attachment

A-8

Section 1EP6: Drill EvaluationProceduresAP 3125 App. A, "EAL Classification Matrix," Rev.22OP 3546, "Operation of the Emergency Operations Facility/Recovery Control," Rev. 30Miscellaneous DocumentsNEI 04-02, "Regulatory Assessment Performance Hot Conditions lndicator Guideline," Rev. 5"January 19,2011 Emergency Preparedness Drill Sequence of Events," Rev. 0Section 4OA1: Performance Indicator (Pl) VerificationCondition ReportsCR-WY-2O1

0-03036, "Automatic reactor scram'CR-WY-201 0-051 28. "Feedwater header leak"Miscellaneous DocumentsESOM-Control Room Narrative Logs January 1,2010 to December 31 ,2010

Section 4OA2: Problem ldentification and ResolutionMiscellaneous DocumentsPassport Reports for current Operator Workarounds, Operator Burdens, and Control RoomDeficienciesAttachment

ADAMSADSAPCAPCFRCRCSDRPDRSEALEDGEPFMEHPCItMcIPEEEISTLORNCVNEINRCOPPARSPIPMTRCrCRCSRHRRHRHXRHRSWSFPCSRASRVsSSCsTPCSTSUFSARA-9

LIST OF ACRONYMS

Agencywide Documents Access and Management SystemAutomatic Depression SystemAdministrative ProcedureCorrective Action ProgramCode of Federal RegulationsCondition ReportCore SprayDivision of Reactor ProjectsDivision of Reactor SafetyEmergency Action LevelEmergency Diesel GeneratorEmergency PreparednessForeign Material ExclusionHigh Pressure Coolant lnjectionInspection Manual ChapterIndividual Plant Examination for External EventsIn-Service TestingLicensed Operator RequalificationNon-cited ViolationNuclear Energy lnstituteNuclear Regulatory CommissionOperating ProcedurePublicly Available Records SystemPerformance lndicatorPost Maintenance TestingReactor Core lsolation CoolingReactor Coolant SystemResidual Heat RemovalResidual Heat Removal Heat ExchangerResidual Heal Removal Service WaterSpent Fuel CoolingSenior Reactor AnalystSteam Safety System Relief ValvesStructures, Systems and ComponentsPower Conversion SystemTechnical SpecificationUpdated Final Safety Analysis ReportVermont YankeeWork OrderVYWOAttachment