IR 05000395/2006007

IR 05000395-06-007; South Carolina Electric & Gas Company; 03/17/2006 - 03/31/2006; Virgil C. Summer Nuclear Station

ML061180276
Person / Time
Site:	Summer
Issue date:	04/28/2006
From:	Landis K NRC/RGN-II/DRP/RPB5
To:	Archie J South Carolina Electric & Gas Co
References
IR-06-007
Download: ML061180276 (29)

Text

April 28, 2006South Carolina Electric & Gas CompanyATTN: Mr. Jeffrey B. ArchieVice President, Nuclear OperationsVirgil C. Summer Nuclear Station

P. O. Box 88 Jenkinsville, SC 29065SUBJECT:VIRGIL C. SUMMER NUCLEAR STATION - NRC PROBLEM IDENTIFICATIONAND RESOLUTION INSPECTION REPORT 05000395/2006007

Dear Mr. Archie:

On March 31, 2006, the United States Nuclear Regulatory Commission (NRC) completed aninspection at your Virgil C. Summer Nuclear Station. The enclosed report documents theinspection findings which were discussed on March 31, 2006, with you and other members of your staff.The inspection was an examination of activities conducted under your license as they relate tothe identification and resolution of problems, compliance with the Commission's rules and regulations, and the conditions of your operating license. Within these areas, the inspection involved selected examination of procedures and representative records, observations of activities, and interviews with personnel.On the basis of the samples selected for review, the team concluded that generally problemswere properly identified, evaluated, and corrected. One finding of very low safety significance (Green) was identified during this inspection associated with repairing Main Steam Isolation Valves. The finding was determined to involve a violation of NRC requirements and has beenentered into your corrective action program. If you contest the non-cited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Virgil C. Summer Nuclear Station.

SCE&G2In 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) com ponent ofNRC's document system ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/Kerry D. Landis, ChiefReactor Projects Branch 5 Division of Reactor ProjectsDocket No: 50-395License No: NPF-12

Enclosure:

Inspection Report 05000395/2006007w/Attachment: Supplemental Information

REGION IIDocket No.:50-395 License No.:NPF-12 Report No.:05000395/2006007 Licensee:South Carolina Electric & Gas (SCE&G) Company Facility:Virgil C. Summer Nuclear Station Location:P. O. Box 88Jenkinsville, SC 29065Dates:March 13-17 and March 27-31, 2006 Inspectors:R. Hagar, Senior Resident Inspector, Robinson (Team Leader)N. Garrett, Senior Resident Inspector, Surry R. Carrion, Project Engineer, RII M. Cain, Resident Inspector, V.C. SummerApproved by:K. Landis, Chief, Reactor Projects Branch 5Division of Reactor ProjectsAttachment:Supplemental Information EnclosureSUMMARY OF ISSUESIR 05000395/2006007; 03/13-17/2006, 03/27-31/2006; Virgil C. Summer Nuclear Station;Biennial baseline inspection of the problem identification and resolution program.The inspection was conducted by two Senior Resident Inspectors, a Project Engineer, and aResident Inspector. One Green finding, which was a non-cited violation, was identified. Thesignificance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC'sprogram for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.Identification and Resolution of ProblemsOverall, the licensee maintained an effective program for the identification and correction ofconditions adverse to quality. However, during the inspection, several minor problems were identified. These problems included two cases in which identification of the problem was not complete, accurate, and timely; three cases in which extent-of-condition was not considered; four cases in which corrective actions that were not appropriately focused to correct the problem; and three cases in which corrective actions were not completed in a timely manner commensurate with the safety significance of the issues. The licensee was generally effective at identifying problems at a low threshold and entering them into the Corrective Action Program (CAP). The licensee consistently prioritized issues in accordance with their CAP and routinely performed adequate evaluations that were technically accurate and of sufficient depth. Rootcause analyses were performed when appropriate and problem evaluations considered extent of condition and generic implications appropriately. Corrective actions were effective in correcting problems. Management fostered a safety-conscious work environment by emphasizing safe operations and encouraging problem reporting.A.NRC-Identified and Self-Revealing Findings Cornerstone: Initiating EventsGreen. A self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V,Instructions, Procedures, and Drawings, was identified for failure to establish, implement, and maintain adequate maintenance procedures to ensure that the Main Steam Isolation Valves (MSIVs) were capable of performing their safety-related function.

Specifically, maintenance procedure MMP-300.023, Main Steam Isolation Valve Air Actuator Maintenance, was inadequate in that it did not include hot-condition c hecks ofthe alignment of the bottom spring plate and stanchion gap tolerances.This finding is greater than minor because it impacts the equipment performanceattribute of the Reactor Safety Mitigating Systems Cornerstone in that the failure of the MSIV to close affects the reliability and availability of that valve. This finding wasdetermined to be of very low safety significance because the valve did go closed within a relatively short time, and because the effects of the failure of a single MSIV to close are bounded by accident analysis assumptions. (Section 4OA2 c.(2).2)B.Licensee-Identified ViolationNone EnclosureReport Details4.OTHER ACTIVITIES4OA2Identification and Resolution of Problems a.Effectiveness of Problem Identification (1)Inspection ScopeThe team reviewed items selected across the three strategic performance areas(reactor safety, radiation safety, and physical protection) to verify that problems were being properly identified, appropriately characterized, and entered into the Corrective Action Program (CAP) for evaluation and resolution. The team reviewed program documents including Revision 0 of Station Administrative Procedure (SAP)-999, Corrective Action Program, which described the administrative process for documenting and resolving problems. The team also reviewed other program documents including SAP-1356, Cause Determination, Revision 1, and SAP-1351, Operating ExperienceProgram, Revision 4.The team attended the licensee's daily management review team meetings March 13 -16 and March 27 - 30, to gauge the effectiveness of the screening process in ensuring that problems were properly captured in the licensee's Condition Evaluation Report (CER) database. The team also listened to daily CER review team phone calls on March 13 - 16, March 27, March 28, and March 30, to determine whether identifiedproblems were properly characterized and prioritized.The team reviewed a sampling of CERs that had been generated or closed since May2004, and focused on CERs associated with systems that ranked high on the licensee'srisk-significance list. For the team, the licensee conducted several computer database searches and sorts to identify the specific attributes associated with the issues identified and documented in the CAP. Those sorts included CERs issued during the past two refueling outages, CERs that described failures of in-service or Technical Specification (TS) surveillance tests since May 2004, operating experience items initiated since May 2004, and CERs that described the evaluations of and corrective actions taken for NRC-identified findings since May 2004.The team reviewed plant equipment issues associated with maintenance rule (a)(1)items, functional failures, maintenance preventable functional failures (MPFFs), and repetitive MPFFs, to verify that maintenance rule equipment deficiencies were being appropriately entered into the CAP.The team toured the plant, including portions of the intermediate building, the auxiliarybuilding, the service water pumphouse, the control room, and the diesel generatorbuilding to determine whether equipment and material condition problems were being identified.

4EnclosureWhile in the control room, the team reviewed the equipment removal and restorationlogbook (all open items), and the logbook of open control room discrepancies to determine if problems potentially affecting safe plant operations were properly entered into the CAP process.In addition, the team reviewed the Nonconformance Control and Corrective ActionsAudit dated 3/2/05, the Corrective Action Program Self-Assessment dated October 11-14, 2004, and the CERs generated as a result of these audits, which included CERs 0-C-06-0090, 0-C-05-0457, 0-C-05-0458, and 0-C-05-0459. The team evaluated theassessment's effectiveness in identifying problems in the CAP process and compared the results of the licensee's efforts with the teams findings and observations.The team reviewed the industry OE program through review of SAP-1351, Revision 4,Operating Experience Program, and interviews with key personnel in the Organization Development & Performance (OD&P) and Nuclear Licensing (NL) departments. Several

NRC generic communications were selected to determine if the licensee had screenedthese items into the CAP by documenting them with a CER. In addition, general review of the CER documentation was performed for selected CERs.Documents reviewed are listed in the Attachment. (2) AssessmentThe team concluded that site personnel were appropriately generating CERs asrequired by the licensee's program.From the review of CERs associated with maintenance rule items and previously issuednon-cited violations (NCVs), the team determined that site personnel were appropriatelydocumenting maintenance rule problems in the licensee's CAP process. Maintenance rule evaluations performed using attachments in procedure ES-514, Maintenance Rule Implementation, were appropriately attached to the associated CER in the Problem Identification Program (PIP) database.The team identified no significant differences between the corrective action processsamples reviewed during this inspection and the licensee's self-assessments and audits that were related to the effectiveness of the licensee's corrective action process.

Therefore, the team concluded that the self-assessments and audits performed by thelicensee were effective in identifying issues, and that deficiencies were appropriately entered into the corrective action process.The team noted that SAP-1351, Operating Experience Program, Revision 4, sufficientlydetailed a process to screen industry operating experience (OE). Industry OE items were routinely reviewed by the licensee's OD&P department and the dispositions of the items were recorded in an OE log.

5EnclosureThe team identified the following two cases in which identification of the problem wasnot complete and accurate:*When CER 0-C-04-790 was initiated on 3/19/04 for Maintenance Rule trackingpurposes to provide the 60-day completion of the functional failure evaluation of the failure of a level switch that had occurred on 1/21/04, the licensee did not identify thedelay in initiating the CER as a problem.*When CER 0-C-04-791 was initiated on 3/19/04 for Maintenance Rule trackingpurposes to provide the 60-day completion of the functional failure evaluation of the failure of a level switch that had occurred on 1/28/04, the licensee did not identify thedelay in initiating the CER as a problem.For these same two cases, the team considered that identification of the problem wasnot timely, because:*CER 0-C-04-790 was initiated on 3/19/04 for a failure that had occurred on 1/21/04.*CER 0-C-04-791 was initiated on 3/19/04 for a failure that had occurred on 1/28/04. b.Prioritization and Evaluation of Issues (1)Inspection ScopeThe team reviewed a sample of corrective action documents to determine if the licenseeappropriately prioritized and evaluated various issues being entered into the CAP. A sample of corrective action documents were selected from the various cornerstones with a focus on issues related to higher risk significant plant systems.The team reviewed selected CERs, including those associated with industry operatingexperience issues and nonconformance notices, to determine whether site personnel conducted reviews for generic implications, repetitive conditions, and common-cause failure mode determinations when the condition warranted.The team attended the licensee's daily management review team meetings March 13 -16 and March 27 - 30, to determine the level of management attention that problemsreceived. The team also listened to daily CER review team phone calls on March 13 -

16, March 27, March 28, and March 30, to determine whether identified problems wereproperly characterized and prioritized.Documents reviewed are listed in the Attachment. (2)AssessmentThe team determined that the licensee was effective in prioritizing and evaluating issuescommensurate with their safety significance.

6EnclosureThe team concluded that the licensee's problem evaluations appropriately consideredextent of condition and generic implications, and operability and reportability of issueswere appropriately evaluated and resolved. At the various management meetings, the team observed that the specific issues identified in CERs received a level of discussion commensurate with their safety significance. The team also concluded that root causeanalyses were being performed when appropriate.The team identified the following three cases in which extent of condition was notconsidered:*CER 0-C-05-4593 described evaluation of an operating-experience item that showedthat failure of a potential transformer in a voltage regulator could cause an over-voltage condition that could affect safety-related components connected to the affected bus. In this evaluation, the licensee did not evaluate the vulnerability of safety-related equipment to the effects of a similar failure at the V.C. Summer station.*CERs 0-C-05-4301 and 0-C-05-1997 described evaluation of the discovery of mudand sludge coming out of a sample valve in the cooling system of an emergencydiesel generator. In this evaluation, the licensee's historical search of related issues failed to identify one previous occurrence.*CER 0-C-05-4440 described evaluation of an operating experience item regarding aphenomenon called "tin whiskering" in safety-related circuits. In this case, the licensee took no action to assess the possible effects of tin whiskering in systemsother than the solid-state protection system, despite vendor recommendations toassess all applications that could be vulnerable to the phenomenon.In addition, the team identified the following two cases in which root and contributingcauses were not identified:*CER 0-C-05-3144 describes an issue related to returning safety-related valves toservice following diagnostic testing with unacceptable results. However, this CER did not describe an evaluation and did not identify any cause; instead, the evaluationsection of the CER described a proposed corrective action, but provided no related basis.*CER 0-C-05-0457 describes an audit-identified finding of a weakness in thelicensee's root-cause analysis process. The evaluation section of this CER did not identify any root or contributing causes. c.Effectiveness of Corrective Actions (1)Inspection ScopeThe team reviewed the CERs listed in the Attachment to verify that the licensee hadidentified and implemented corrective actions commensurate with the safety significance 7Enclosureof the documented issues, and where possible, evaluated the effectiveness of theactions taken. The team also verified that common causes and generic concerns were addressed where appropriate. In addition, the teams reviewed CERs associated with previous NCVs to assess the adequacy of corrective actions.Documents reviewed are listed in the Attachment. (2)Assessment .1General ObservationsFrom the CER reviews, the team determined that the licensee's corrective actions wereeffective in correcting problems. Management involvement in the Corrective Action Review Board (CARB) process was also considered to be effective. The team also concluded that corrective actions for previous NCVs were adequate.The team identified the following four corrective actions as not appropriately focused tocorrect the problem:*In CER 0-C-06-0090, sequence 1, the corrective action to improve properapplication of CER action level codes was to discuss the proper application of CER action levels and CER categories with the Unit Evaluators and Management Review Team members at the combined CER review team and Management Review Team meeting held on February 8, 2006. The inspection team considered this corrective action to be not appropriately focused to correct the problem because it was a one-time action that was not made part of a process that would provide periodic reinforcement and accountability.*In CER 0-C-06-0090, sequence 1, the corrective action to improve departmentaltrending efforts was for the site Vice President task departmental managers to provide departmental trends for their areas at the next Plant Challenge meeting. The inspection team considered this corrective action to be not appropriately focused to correct the problem because it was a one-time action that was not made part of a process that would provide periodic reinforcement and accountability.*In CER 0-C-06-0090, sequence 1, the corrective action to improve assignment ofunit evaluators was to use the teleconference conference system for the CER review team meetings and for unit evaluators to receive reinforcement from senior management. The inspection team considered this corrective action to be not appropriately focused to correct the problem because it was not made part of a process that would provide periodic reinforcement and accountability.*In CER 0-C-05-4440, the licensee's corrective actions were not appropriatelyfocused to correct the problem of tin whiskering in safety-related circuits because the licensee did not initiate a preventive-maintenance task to clean and inspect installed circuit cards and in-stock spares on a regular basis, despite a vendor 8Enclosurerecommendation to do so. In this case, the licensee deviated from a vendor-recommended practice with no documented basis.In addition, the team identified the following three cases in which corrective actions werenot completed in a timely manner commensurate with the safety significance of the issue:*CERs 0-C-01-0403 and 0-C-04-0741 described a nonconforming condition in whicha pipe clamp on a pipe in the component cooling water system (a safety-relatedsystem) was making contact with a nearby piece of structural steel. This issue wasidentified in CER 0-C-01-0403 on 3/21/01 and was resolved under CER 0-C-04-0741 on 5/11/05. In the process, CER 0-C-01-0403 remained open for 1111 days, and CER 0-C-04-0741 remained open 540 days. Disregarding the overlap between CER 0-C-01-0403 and CER 0-C-04-0741, this nonconforming condition was corrected 1512 days after it was first identified.*In CER 0-C-05-4400, the evaluation of tin whiskering in safety-related circuits wascompleted approximately ten months after the related operating experience reportwas received.*In CER 0-C-04-1064, the corrective action to replace components used to maintainthe temperature of the diesel-driven fire pump warm during cold environmental conditions was completed 473 days after the condition was identified. .2Inadequate Maintenance Procedures to Repair Main Steam Isolation ValvesIntroduction: A self-revealing Green non-cited violation (NCV) was identified for failingto establish, implement, and maintain adequate maintenance procedures to ensure that the Main Steam Isolation Valves (MSIVs) were capable of performing their safety-related function. Specifically, maintenance procedure MMP-300.023, Main Steam Isolation Valve Air Actuator Maintenance, was inadequate in that it did not include hot-conditionchecks of the alignment of the bottom spring plate and stanchion gap tolerances. Thisinadequacy resulted in the failure of XVM02801B-MS, 'B' Main Steam Isolation Valve, to go fully closed.Description: On December 7, 2004, in Mode 2 during a forced plant shutdown due to asteam line break on the high-pressure main turbine, MSIV XVM02801B-MS failed to fully close the first five times that control room operators attempted to close the valve fromthe Main Control Board (MCB). Those attempts resulted in the valve indicating 'mid-position' both locally and on the MCB. On the sixth attempt, the valve went fully closed.

Subsequent investigation and root cause analysis determined that misalignment of thebottom spring plate with the yoke (stanchion) rods had caused increased friction and binding which prevented the valve from fully closing. The misalignment was attributed to thermal expansion tolerance changes and the wear of 'brass tipped' adjustment pins which hold the bottom spring plate away from the stanchion rods. The licensee's corrective action to prevent recurrence was to initiate a preventive-maintenance task sheet to ensure the alignment of the bottom spring plate and stanchion gap tolerances 9Enclosureof the valve during hot conditions coming out of a refueling outage or after maintenance. The licensee also enhanced maintenance procedure MMP-300.023, Main Steam Isolation Valve Air Actuator Maintenance, to include guidance for conducting hot-condition tolerance and set-up of the bottom spring plate and stanchions.Analysis: The inoperability of valve XVM02801B-MS resulted from the performancedeficiency of not including adequate instructions in procedure MMP-300.023. The finding is considered greater than minor because it had a direct impact on the MSIV to perform its safety-related function, which is to close during a high-energy line break or steam generator tube rupture. This finding impacts the equipment performance attribute of the Reactor Safety Mitigating Systems Cornerstone in that the failure of the MSIV to close affects the reliability and availability of that valve. This finding wasdetermined to be of very low safety significance because the valve did go closed on the sixth attempt, and failure of a single MSIV to close is within the accident analysis assumptions.Enforcement: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures andDrawings, requires that all activities affecting quality be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Criterion V further requires that instructions, procedures, or drawings include appropriate quantitative or qualitative acceptance criteria for determining that importantactivities have been satisfactorily accomplished. Contrary to the above, the licensee failed to establish, implement, and maintain adequate maintenance procedures to ensure that the Main Steam Isolation Valves (MSIVs) were capable of performing their safety-related function, in that maintenance procedure MMP-300.023, Main SteamIsolation Valve Air Actuator Maintenance, was inadequate because it did not include hot-condition checks of the alignment of the bottom spring plate and stanchion gaptolerances. Because this finding was determined to be of very low safety significance (Green) and the licensee has entered this issue into their corrective action program as CER 0-C-06-1110, this violation is being treated as a non-cited violation (NCV),

consistent with Section VI.A of the NRC's Enforcement Policy, and has been designatedNCV 06000395/2006007-01, Failure to Use Adequate Maintenance Procedures to Inspect and Repair Main Steam Isolation Valves. d.Assessment of Safety-Conscious Work Environment (1)Inspection ScopeThe team informally interviewed licensee personnel to develop a general view of thesafety-conscious work environment and to determine if any conditions existed that would cause workers to be reluctant to raise safety concerns. The team also discussed issues with the current Senior Resident Inspector (assigned since 2004) to gain his perspective on the site safety-conscious work environment. The team also reviewed the licensee's employee concerns program (ECP), which provides an alternate method to the CER process for employees to raise safety concerns with the option of remaining anonymous.

10Enclosure (2)AssessmentThe team concluded that licensee management generally fostered a safety-consciouswork environment by emphasizing safe operations and encouraging problem reporting.

Methods available to encourage problem reporting included CERs, maintenance work requests, and the ECP. Interviews with licensee staff did not identify any reluctance to report safety concerns.4OA6MeetingsExit Meeting SummaryThe team presented the inspection results to Mr. and other members oflicensee management at the conclusion of the inspection on March 31, 2006.The team asked the licensee whether any of the material examined during theinspection should be considered proprietary. No proprietary information was identified.ATTACHMENT: SUPPLEMENTAL INFORMATION AttachmentSUPPLEMENTAL INFORMATIONPARTIAL LIST OF PERSONS CONTACTEDLicenseeL. Bennett, Plant Support Engineering SupervisorM. Carr, Plant Support Engineering Engineer L. Cartin, Design Engineering Engineer T. Clark, Plant Support Engineering Supervisor A. Cribb, Nuclear Licensing Supervisor G. Croxton, Plant Support Engineering Engineer/BS K. Culp, Organizational Development & Performance Specialist D.

Deardorff,

Sr. Engineer/CAP D. Dobson, Plant Support Engineering Engineer J. Garza, Plant Support Engineering Engineer L. Harris, Plant Support Engineering Engineer J. Heilman, Organizational Development & Performance Supervisor/CAP M. Hicks, Plant Support Engineering EngineerD. Jones, Design Engineering Engineer D. Lavigne, General Manager, Organizational Effectiveness G. Lippard, Operations Manager K. Marsh, Organizational Development & Performance Specialist T. Matlosz, Organizational Development & Performance Manager D. McGlauflin, Organizational Development & Performance Specialist F. McKinnon, Plant Support Engineering Engineer F. O'Neal, Plant Support Engineering Engineer V. Pearson, Organizational Development & Performance Specialist/CAP G. Robertson, Plant Support Engineering Engineer B. Schwartz, Maintenance Rule Coordinator W. Stuart, Plant Support Engineering Manager M. Torres, Plant Support Engineering Engineer J. Turkett, Nuclear Licensing Engineer B. Waldrop, Plant Support Engineering Engineer T. Welsh, Plant Support Engineering Engineer R. Word, Plant Support Engineering Supervisor NRCJohn Zeiler, Senior Resident Inspector A-2AttachmentITEMS OPENED, CLOSED AND DISCUSSED Open/Closed05000395/2006007-01NCVFailure to Use Adequate Maintenance Procedures toInspect and Repair Main Steam Isolation Valves (Section 4OA2 c.(2).2)DiscussedNone A-3AttachmentLIST OF DOCUMENTS REVIEWEDCondition Evaluation Reports0-C-98-0754, Change TS section 4.5.2.b.2 to delete the reference to ECCS pump casings.0-C-98-1047, Groundwater in-leakage at the (Residual Heat Removal) and (Reactor Building)spray pipe penetrations through the AB wall.0-C-99-0007, The A Condensate Pump discharge valve took 8 minutes to open during a planttransient.0-C-99-0131, PZR (Pressurizer) Backup Group 2 Heater SWGR (Switchgear) tripped withoutapparent reason.0-C-02-0916, Groundwater in-leakage around guard pipes and coating/concrete failure aroundpenetration sleeves.0-C-02-1726, The B Station Instrument Air Compressor was given a start signal and would notload.0-C-02-1757, A Diesel Generator breaker tripped open while paralleled to offsite power.

0-C-02-1980, Air header pressure started dropping with B Instrument Air compressor running at85 psig; manually started A compressor which immediately tripped on high air temperature; started supplemental air compressor and recovered air pressure.0-C-02-2500, Switch did not actuate during functional test (ILS01929).

0-C-02-2748, A Instrument Air Compressor tripped on low oil pressure on startup.

0-C-02-2820, TS Amendment: CER created to transfer this issue from Licensing PIP 0-L-99-0138. [Licensee Event Report] 1998-008: Missed [Technical Specification] SurveillanceRequirement to vent [residual heat removal] pump casings. Opened to track completion of

[Licensee Event Report], Tech Spec amendment & NRC Inspection items. 0-C-02-2872, The security latch on [door] DRCB/503 failed.0-C-02-2883, Field failure of 40DG relay caused [Diesel Generator] A output breaker to open.

0-C-02-3061, Instrument Air Compressor B surge alarm and inlet filter high differential pressurereceived; compressor would not maintain system pressure.0-C-02-3688, A [Emergency Diesel Generator] experienced high lube oil strainer [differentialpressure] during the surveillance run per [Surveillance Test Procedure] -125.002A, STTS#0218392.0-C-03-0433, RIS 2003-02: Importance of giving NRC advance notice of license renewal.0-C-03-0506, [V.C. Summer Nuclear Station] [Nuclear Regulatory Commission] IntegratedInspection Report # 50-395/02-04. Four (4) non-cited violations of low significance (Green)

were identified during the inspection period of September 29, 2002 through January 4, 2003.0-C-03-1847, Corrective Action Program does not have an effective process for identifyingrepetitive events. 0-C-03-2171, Lube Oil Strainer [differential pressure] exceeded log limits during surveillancetesting.0-C-03-2837, XVC03162A failed leakage testing during the performance of [surveillance testprocedure] 123.003A.0-C-03-3364, Wires inadvertently loosened while performing relay change-out.

0-C-03-3500, During run-in on A Diesel Generator, the fuel supply line on the right bankruptured at the threads.0-C-03-3663, Lead found not connected on one side of diode D1 in XPN5504 [DieselGenerator] B Control Panel.

A-4Attachment0-C-03-3800, Sample valve XVT-20989-DG for sampling B Diesel Generator Jacket CoolingWater appears to be clogged.0-C-03-4348, Unable to control speed of turbine driven emergency feedwater pump duringsurveillance test.0-C-03-4399, While performing [preventive maintenance tasking sheet] #0313350 ILS01902 didnot actuate.0-C-04-0049, QA-AUD-200312-0 Station does not adequately recognize and classify risksignificance relative to the [condition evaluation report] Process. The [corrective action program] lacks organizational ownership, upper management support, and la cks anadequate corrective action follow-up process.0-C-04-0165, Westinghouse Technical Bulletin, TB-04-2, Solid State Protection System MasterRelay, that needs to be reviewed for applicability to VCSNS.0-C-04-0187, New Federal Signal Electronic Sirens may have defective amplifiers that could beadversely affected by cold temperatures. These sirens have not yet been installed.0-C-04-0188, OE17156 - [Solid State Protection System] Safeguard Driver Board degradationfound during inspections. 0-C-04-0223, Ice storm resulted in the loss of greater than 25% of the Early Warning SirenSystem.0-C-04-0241, OE17609 - Reactivity Excursion upon Starting Standby Centrifugal ChargingPump for Surveillance Test.0-C-04-0289, NRC questioned gap on pivot point for trip throttle valve, XVT02865-MS on theTurbine Driven Emergency Feedwater Pump.0-C-04-0355, Pressurizer Back-up Group 2 Heaters tripped again for no apparent reason.

0-C-04-0389, The audible alarm for the [Early Warning Siren System] computer was foundsilenced which disabled the prompting that the [Early Warning Siren System] operability haddropped below 85%.0-C-04-0460, NRC Information Notice 2004-01: Auxiliary Feedwater Pump Recirculation LineOrifice Fouling - Potential Common Cause Failure. NRC document that needs to bereviewed for applicability.0-C-04-0462, The station encountered a loss of greater than 25% of the EWSS capability dueto an ice storm which interrupted electrical power to nearly half of the sirens0-C-04-0479, This [condition evaluation report] is being initiated to provide a mechanism totrack any procedural corrective actions which may be determined to be required relative to previously closed CER-03-0128.0-C-04-0747, While placing XFL-8A in service, Reactor Coolant Pump Seal Injection flows on Aand B [Reactor Coolant Pumps] decreased to 2.7 and 4.1 gpm respectively.0-C-04-0766 ,The corrective action to resolve the pipe support condition in CER-01-0403 isconsidered untimely. CER-04-0741 should address if an NCN condition existed since thecondition was identified in 3/21/03 (refer to CER-01-0403).0-C-04-0790, The [condition evaluation report] is generated for Maintenance Rule trackingpurposes to provide the 60-day completion of the functional failure evaluation of level detection system Level Switch ILS01926.0-C-04-0791, The [condition evaluation report] is generated for Maintenance Rule trackingpurposes to provide the 60-day completion of the functional failure evaluation of level detection system Level Switch ILS01905.

A-5Attachment0-C-04-0818, OE17978 - Foreign Material Found in Component Cooling Water HeatExchanger. This is an Operating Experience issue that needs to be evaluated for applicability to VCSNS.0-C-04-0879, Apparent [reactor coolant system] boundary leakage identified at the sealinjection line to the C [reactor coolant pump] during the reactor building building entry today while investigating unidentified [reactor coolant system] leakage.0-C-04-0881, [reactor coolant system] boundary leakage identified at the seal injection line tothe C [reactor coolant pump]. This [condition evaluation report] will document issues relatedto Boric Acid and the necessary evaluations per SAP-1100.0-C-04-0884, [Licensee Event Report] 2004-001: [valve] IFV00498-FW apparently failed closedwhile decreasing power resulting in a reactor trip due to Lo-Lo steam generator (S/G) in the C S/G.0-C-04-0988, This [condition evaluation report] is to track the Root Cause Analysis on the CReactor Coolant Pump Seal Injection Line leak and associated corrective actions.0-C-04-1064, Diesel Fire Pump, XPP0134B, Keep Warm System not maintaining temperaturewarm.0-C-04-1069, ILT00459 failed low while attempting to electrically isolate IPT00950 on W.O.

0407763.0-C-04-1121, Reactor Coolant System is leaking into Residual Heat Removal discharge headerand level control valve 115A. Reactor Coolant System Unidentified leakage is 0.23 gpm with flow through letdown demins. Leakage is 0.13 gpm with demins bypassed.0-C-04-1239, Steam Propagation Door DRCB/302 found open and unattended.0-C-04-1385, Pressurizer (PZR) Backup Group 2 Heater breaker spuriously tripped.

0-C-04-1490, During SSPS A train logic testing, Slave A test pushbutton (position 18) did notrespond as required. A4 LED was found on prior to testing that position and would not extinguish.0-C-04-1527, RCP seal leakage in accordance with WCAP-10541 may no longer be acceptableto the NRC for use in the FPER and its supporting Safe Shutdown (SSD) analysis.0-C-04-1786, Lube oil leak on A diesel generator shaft driven pump while engine in operation.

0-C-04-1900, RCS Total Flow Rate is decreasing. If trend continues, A loop flow rate willdecrease below the 100% loop design flow rate.0-C-04-2170, B Feedwater Pump oil coolers appear to be fouled or fouling based ontemperature rise in the oil operating temp.0-C-04-2262, Plant downpower due to high circulating water (CW) temp.

0-C-04-2308, Operating Event (OE)18617 - Main Feedwater Regulating Valve (MFRV) Plug-Stem Separation (Follow-Up to OE 17781), is an OE event that needs to be evaluated for applicability to VCSNS.0-C-04-2381, NRC Integrated Inspection Report 05000395/2004003 identified two Greenfindings Non-Cited Violation (NCV) on Reactor Coolant Pump (RCP) C weld repair and NCV on drawing and NCV on drawing control loss of Reactor Coolant System (RCS) Pressurizer heater control.0-C-04-2388, Addresses the Green licensee-identified non-cited violation on the disabledcomputer alarm associated with the EWSS documented in NRC Report 05000395/2004002.0-C-04-2410, Sight glass/indicator for feedwater heater is obsolete.

0-C-04-3051, Box purchased through Investment Recovery was found to contain packagedradioactive materials.0-C-04-3121, Operating Experience Screening Self-Assessment SA04-OD-05.

A-6Attachment0-C-04-3230, While performing a maintenance history review for several level switches in theLD (Level Detection) system, found three level switches that had failed functional testing, butnot declared as Maintenance Rule Functional Failures.0-C-04-3262, A Residual Heat Removal (RHR) pump (XPP0031A) breaker (XSW1DA1 06A)failed to close during attempted start per STP 205.004.0-C-04-3386, [Licensee Event Report] 2004-003: During the restart of the A train loadingsequencer after completion of the maintenance run on A diesel generator (DG) the undervoltage test switch was inadvertently actuated.0-C-04-3644, Corrective Action Program Self-Assessment SA-04-OD-02.

0-C-04-3702, Damper XDP0023B open limit switch did not make up during an attempted fan(XFN0030) start.0-C-04-3755, QA Audit QA-AUD-200414-0, Nonconformance Control, QA finding. Programinadequacies and violations were noted regarding the issuance and control of QA Findings via PIP [condition evaluation report] database. 0-C-04-3772, During the unit shutdown, the C feedwater regulating valve (FWRV) responsewas very slow.0-C-04-3775, Extraction Steam Pipe Failure.

0-C-04-3786, B Main Steam Isolation Valve had mid position indication when valve was closed.

0-C-04-3803, IPT00950 Indication Drifting Low.

0-C-04-3840, Quality Service Observation of activities associated with [condition evaluationreport] 04-3803 (IPT-0950).0-C-04-3856, C SW (Service Water) Pump breaker (XSW1EB 02) charging springs failed tocharge.0-C-04-3931, XSWDB-05, PZR (Pressurizer) Backup Group 2 Heater breaker tripped.

0-C-04-4366, PZR (Pressurizer) Backup Group 2 Heater breaker tripped again for no apparentreason.0-C-05-0069, XSWDB-05, PZR (Pressurizer) Backup Group 2 Heater breaker tripped duringmanual makeup.0-C-05-0142, A diesel generator (DG) air start valve was leaking air when XVP10987A-DG wasopen during return to operation (RTO) clearing. While investigating the source of air leakage the A DG started in Maintenance and was secured. The RPMs exceeded 115 but the DG never reached operating speeds.0-C-05-0230, OE19460 - Time Delay Relays Exceed TS Allowable Value.

0-C-05-0299, NRC (Nuclear Regulatory Commission) Inspection Report 2004-005 covers athree-month period (9/26-12/31/04).0-C-05-0306, IFV00478, 488, 498 Maintenance Recommendations from information obtained atthe 2005 Air-Operated Valve (AOV) users group conference.0-C-05-0330, The required documentation for a Maintenance Rule FunctionalFailure/Exceeding Performance Criteria following event declaration was not completed within the time prescribed in ES-0514.0-C-05-0402, OE19880 - Multiple TGSCC Found in RCS Cold Leg Sample System Tubing(Beaver Valley 1).0-C-05-0457, QA Audit QA-AUD-200414-0 identified a finding in the root cause analysisprocess.0-C-05-0458, QA Audit QA-AUD-200414-0 identified that corrective action levels were assignedincorrectly per SAP-1131, Enclosure B.

A-7Attachment0-C-05-0459, During audit QA-AUD-200414-0 identified Finding regarding SAP-1142, Trendingof Station Deficiencies.0-C-05-0573, While testing B Diesel Generator (DG) annunciator, the IB operator notified theControl Room that he was hearing relay chatter and auxillary equipment was starting andstopping.0-C-05-0616, OE19872 - Improper linkage adjustment on moisture separator level switchescaused main turbine trip. - Catawba Nuclear Station.0-C-05-0731, The replacement Reactor Coolant Pump studs to be installed during RF-15 didnot addressSection XI requirements for baseline Ultrasonic Inspection.0-C-05-0847, OE20033 - Update to OE19734 - Diesel Generator Speed And Voltage Failed ToReach Required Values In Under 10 Seconds.0-C-05-0856, Damper limit switch on XDP0023B in need of adjustment 0-C-05-0975, Measured outside air flow on A control room ventilation emergency mode to be1029 scfm.0-C-05-0977, Apparent missed surveillance of both trains of CB ventilation outside air inleakagemeasurement.0-C-05-0993, High differential pressure noted on B D/G (Diesel Generator) lube oil duplexstrainer.0-C-05-1101, VCS (V.C. Summer) does not currently have a procedure to address multipleloss of annunciators.0-C-05-1282, Recommend reactor engineer (RE) revise appropriate reactor engineeringprocedures (REPs) to remove manipulation of valve 8484 and that operating procedures (OPs) revise appropriate station operation procedures (SOPs) to lower volume control tank (VCT) level during approach-to-criticality rather than manipulate valve 8484 which could introduce gas into the vent ed system.0-C-05-1412, Problems occurred when loading diesel air compressor (XAC0014)

0-C-05-1505, During outage work activities on the emergency feedwater turbine (EFW) Turbine(TPP0008), engineering noted as-built differences on the stem drain lines for the Trip Throttle valve and the Governor valve.0-C-05-1508, During outage work activities on the emergency feedwater (EFW) Turbine(TPP0008), engineering noted as-built differences on the stem drain lines for the Trip Throttle valve and the Governor valve.0-C-05-1518, OE 20450-Preliminary-Millstone Unit 3 Automatic Reactor Trip needs to beevaluated.0-C-05-1688, CR Ventilation vendor test results when compared to our system line-up do notappear to be accurate.0-C-05-1729, High lube Oil Strainer (XEG0001B-E) Differential Pressure.

0-C-05-1799, While inspecting SSPS (Solid State Protection System) Universal Logic Card perMillstone OE (Operating Experience), found degraded diode CR47.0-C-05-1809, TB-05-4: During inspection of SSPS (Solid State Protection System) circuitboards per OE 20450, Westinghouse Briefing on SSPS (Solid State Protection System) Tin Whiskers, a tin whisker was identified on a Universal logic card, SN 3113.0-C-05-1891, The ultrasonic thickness measurement for MIC (Microbiologically-InducedCorrosion) component IB-221-12 shows remaining wall thickness less than the established minimum wall. This is downstream of XVC-3125A- SW.0-C-05-1997, When sampling A D/G (Diesel Generator) jacket water cooling, I discovered mudand sludge coming out of sample valve for inside the cooling system.

A-8Attachment0-C-05-2042, [Licensee Event Report] 2005-001: Relay testing per EMP405.024 caused a lossof all balance of plant busses, loss of 1DB, and auto-start of B DG (Diesel Generator).0-C-05-2178, 115-kilovolt (kV) bus voltage below allowed limits. Station Operating Procedure(SOP)-304 limit for current plant configuration is 116.4 kV; actual voltage on bus was 115.1 kV.0-C-05-2196, The A Safety Injection (SI) Accumulator was overfilled due to faulty levelindication caused by four transmitter isolation valves being closed through the valve lineup indicated these valves were open.0-C-05-2226, Approximately 3 gpm leakage into the Pressure Relief Tank (PRT); the reason forthe inleakage for the PRT was determined to be XVR08121: the seal return path relief valve to the PRT.0-C-05-2286, B RHR pump was required to be placed in Pull to Lock (PTL) per GOP-2 at RCStemperature of 250 degrees Fahrenheit (F). It was not placed in PTL as required untilapproximately 258 F.0-C-05-2300, Licensee Event Report (LER) 2005-002: A Motor-Driven Emergency Feed Pump(MDEFP) switch was in PTL , rendering the pump inoperable during Mode 3 escalation.0-C-05-2399, TR5-47 - Review of circuit card/board related failures that contributed toautomatic and manual scrams (addendum to TR5-43), is an OE (Operating Experience)

event that needs to be evaluated.0-C-05-2620, Review of (Refueling Outage)15 (Condition Evaluation Report)s indicates thatthree major human-performance errors occurred in managing plant configuration during (Refueling Outage)15.0-C-05-2728, A DG (Diesel Generator) alarm panel malfunction.

0-C-05-2948, When OE Report 20940 was reviewed, it was noted that the V. C. SummerNuclear Station (VCSNS) had not evaluated Fisher Information Notice 2004-02 and some other earlier vendor notices.0-C-05-3036, Electrical (ES) System Maintenance Rule status goes to (a)(1).

0-C-05-3055, Commenced STP0125.008 (Diesel Generator Load Test) after the completion ofSTP0125.013 (Diesel Gen. A Semi-annual Operability Test). Could not obtain 110% loadper recorder (IYR01803) on XCP6224. Main control board indication reached 4680KW.0-C-05-3144, Test personnel are not recording test parameters and allowable values onICP-240.169 Valve Diagnostic Procedure attachment as required per procedure.0-C-05-3191, Measured stroke time on XVB0003B-AH greater than max allowed on STP-124.001 (510449).0-C-05-3218, Valve found in abnormal condition due to diaphragm leak. With B charging pumpin pull-to-lock, valve XVG09684B-CC is open.0-C-05-3237, While in progress of racking up B Charging pump breaker (XSW1DB 15), breakerspring failed to charge.0-C-05-3238, While in progress of racking up B Charging pump breaker (XSW1DB 15), breakerspring failed to charge. ([condition evaluation report] deleted to 0-0-C-05-3238)0-C-05-3300, This [condition evaluation report] is to document the Maintenance Rule (a)(1) goalcondition on the Instrument Air (IA) System, established on 8/18/05 for exceeding the IAPerformance Criteria 2c, Diesel Air Compressor Unavailability, due to a MaintenancePreventable Functional Failure (MPFF).0-C-05-3329, TDEFW (Turbine Driven Emergency Feedwater) Pump bearing cooling pipingwas noted to have incomplete piping supports.

A-9Attachment0-C-05-3349, [Licensee Event Report] 2005-003: Reactor Trip resulted from failure of ACondensate Pump Discharge Valve, XVB00614A-CO to open while starting A Condensate pump following trip of B Condensate Pump.0-C-05-3357, Westinghouse Technical Bulletin TB-04-22, Revision 1: Reactor Coolant PumpSeal Performance and Appendix R Compliance.0-C-05-3429, OE 20825 - Unplanned scram caused by surveillance test being performed whileanother surveillance test had been temporarily halted for troubleshooting (Browns Ferry) isan OE event that needs to be evaluated for applicability to VCSNS0-C-05-3527, Possible loose part internal to valve actuator.

0-C-05-3621, Feedwater (FW) System IFV00498-FW) Mainteneance Rule status goes to (a)(1)

0-C-05-3753, Sparks due to metal to metal rubbing issued from A CCW (Component CoolingWater) pump inboard seal when pump started for re-test.0-C-05-4301, While sampling the B DG (Diesel Generator) jacket water cooling, mud andsludge was discovered to be coming from the sample tap.0-C-06-0090, Corrective Action Program Self-Assessment SA05-OD-05 0-C-06-0142, Fisher Information Notice: FIN 93-03, Supplement 1, Possible butterfly valvetaper pin failures. Operating experience document that needs to be reviewed for applicability.0-C-06-0393, TSC (Technical Support Center) Engineering Area rearranged without consultingEmergency Services. Labeled Emergency Response Equipment was removed and placed into storage contrary to posted notice.Operating Experience CERs0-C-04-1250, OE18120 - Potential Common Mode Failure for Instrument Air. This is anOperating Experience event that needs to be evaluated for applicability to VCSNS.0-C-04-1413, Westinghouse Infogram IG-04-4 recommends setting the nuclear instrumentation (NI) plateau voltage to 600 Volts vs. the 800 Volts they are now set at or set by a plateau calculation.0-C-04-3457, OE19054 - High Temperature on the Turbine Driven Auxiliary Feedwater Pump(TDAFP) Coupling End Bearing following corrective maintenance.0-C-04-3726, Westinghouse Technical Bulletin TB-04-22: Reactor Coolant Pump SealPerformance and Appendix R Compliance. Operating Experience information to be reviewed for applicability to VCSNS (V.C. Summer Nuclear Station).0-C-05-0956, OE20024 - River Bend (Update to OE19690) Loss of Division 1 Safety BusDuring Surveillance Testing, is an OE event that needs to be evaluated for applicability to VCSummer Nuclear Station (VCSNS).0-C-05-1079, SER2-05 - Gas Intrusion in Safety Systems, is an OE event that needs to beevaluated for applicability to VCSNS.0-C-05-1337, NRC Information Notice (IN) 2005-08: Monitoring vibration to detectcircumferential cracking of reactor coolant pump and reactor recirculation pump shafts.

Operating experience information that needs to be reviewed by VCSNS.0-C-05-2255, NRC Information Notice (IN) 2005-11: Internal flooding/spray-down of safety-related equipment due to unsealed equipment hatch floor plugs and/or blocked floor drains.

Industry Operating Experience that requires VCSNS review.0-C-05-3712, SEN-257 - Internal Flood Design Deficiencies.

0-C-05-3838, OE21175-(Byron)- 2B Feedwater Pump Trip Causes Unit Runback - Update to OE20724.

A-10Attachment0-C-05-4032, SER4-05 - Errors in the Preparation and Implementation of Modifications.0-C-05-4276, TB-05-2 - Potential Shorting of Printed Circuit Board Used in Barton Model 763APressure and Model 764 Differential Pressure Transmitters. Operating Experience document that needs to be reviewed for applicability to VCSNS (V.C. Summer NuclearStation).0-C-05-4345, NRC Information Notice (IN) 2005-30, Safe Shutdown Potentially Challenged byUnanalyzed Internal Flooding Events and Inadequate Design. Industry operating experience document that needs to be reviewed for applicability.0-C-05-4400, NRC Information Notice 2005-25, Inadvertent Reactor Trip and Partial SafetyInjection Actuation Due to Tin Whiskers. Industry operating experience issue that needs to be reviewed for applicability.0-C-05-4593, OE21532 - Trip of the System Auxiliary Transformer (SAT) Feed Breaker to Bus143 Due to Ground Fault in Potential Transformer.Root-Causes AnalysesRC-03-0006, License Amendment Request - LAR 02-2820, Emergency Core Cooling Systems -Exclusion of Safety Injection Pumps from the Requirement to Vent ECCS Pumps.RC-04-0988, Weld Failure and Pressure Boundary Leakage on C Reactor Coolant Pump SealInjection Nozzles.RC-04-1069, Wrong Lead Lifted Due to Use of Illegible Drawing Resulting in Loss of Letdown, 6/3/04.RC-04-3386, Inadvertent Loss of Bus 1DA, 2/2/05.

RC-04-3786, Main Steam Isolation Valve XVM02801B-MS Indicated Mid Position WhenStroked Closed.RC-04-0884, [Licensee Event Report] 2004-001: Feedwater Reg Valve IFV00498.

RC-04-0884, Revision 1, [Licensee Event Report] 2004-001: Feedwater Reg Valve IFV00498.

RC-05-0035, Licensee Event Report 2004-003-01, Safety System Actuation Due to Inadvertentactuation of IDA Undervoltage Test Switch - Supplement 1.RC-05-0069, Pressurizer Backup Heater Breaker Trip During Manual Makeup.

RC-05-0069, Addendum, Pressurizer Backup Heater Breaker Trip During Manual Makeup.

RC-05-0091, Withdrawal of License Amendment Request - LAR 02-2820, Emergency CoreCooling Systems - Exclusion of Safety Injection Pumps from the Requirement to Vent ECCS Pumps.RC-05-0731, The replacement (Reactor Coolant Pump) studs to be installed in RF-15 did notaddressSection XI requirements for baseline Ultrasonic Inspection.RC-05-2042, Loss of the (Balance of Plant) and 1DB Buses during Testing of 86T3 DifferentialLockout Relay, Revision 1.ProceduresES-514, Maintenance Rule Program Implementation, Rev. 3.FPP-025, Fire Containment, Rev. 3.

MMP-300.023, Main Steam Isolation Valve Air Actuator Maintenance, Revs. 2, 3, and 4.SAP-1351, Operating Experience Program, Rev. 4.

SAP-1356, Cause Determination, Rev. 1.

SAP-143, Preventative Maintenance Program, Rev. 11.

SAP-999, Corrective Action Program, Rev. 0.

A-11AttachmentSTP-105.006, Safety Injection/Residual Heat Removal Monthly Flowpath Verification Test,Rev. 11.STP-130.004D, Main Steam Isolation Valve Full Stroke Test (Mode 4), Rev. 1.Other Documents1MS-94B-865-6, Main Steam Isolation Valve Vendor Technical Manual.BWXT Report 1318-001-04-17, Analysis of a Leaking Reactor Coolant Pump (RCP) SealInjection Nozzle Flange to Thermal Barrier Weld Joint From V.C. Summer Nuclear Station, May 2004.Engineering Information Request 81112, [condition evaluation report] 05-3191, XVB00003B-Ahexceeds maximum allowable stroke time of 2 seconds dated 8/14/05.Engineering Change Request 50064, Diesel Generator Air Start After Cooler Removal.

Engineering Change Request 70222, Revise and update effected documents due to identifiedchanges in system pressure or modifications to system.Engineering Change Request 70818, Revised guidance for restoration of RCP (ReactorCoolant Pump) seal cooling during an Appendix R fire.Failure Mechanism Analysis (FMA)04-884.

Information Notice 2005-11: Internal flooding/spray-down of safety-related equipment due tounsealed equipment hatch floor plugs and/or blocked floor drains.Licensee Event Report 1998-008-00, Missed Surveillance Test for ECCS Subsystems - Tavg350FLicensee Event Report 2004-001-00, Reactor Trip Due to Valve Failure During ForcedShutdownLicensee Event Report 2005-002-00, Mode 3 Entry with an Inoperable Emergency FeedwaterPump.Letter, BWXT to Stokes, Evaluation of Leak in C RCP Seal Injection Nozzle, May 4, 2004.

Main Steam [Important to Maintenance Rule] System Function Worksheet dtd 9/12/00.Main Control Room Operator Logs for 12/7/04 - 12/10/04.Maintenance Rule (a)(1) System Goal Status, February, 2006.

Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir Supply, dated 07/03/02.Maintenance Rule Expert Panel Meeting Minutes for 7/14/05.

Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir Supply, dated 09/21/05.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir Supply, dated 08/22/05.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir System, dated 03/03/05.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: EmergencyFeedwater System, dated 10/27/05.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir Supply, dated 06/25/02.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: InstrumentAir System, dated 10/10/02.Maintenance Rule Unacceptable Performance of Failure Cause Determine Form: ResidualHeat Removal System, dated 02/01/05.Maintenance Rule Expert Panel Meeting Minutes for January 4, 2006.

A-12AttachmentMaintenance Rule Expert Panel Meeting Minutes for July 15, 2005.Maintenance Rule Expert Panel Meeting Minutes for August 2, 2004.

Plant Health Committee Meeting Minutes for Thursday, March 2, 2006.

Plant Health Committee Meeting Minutes for Tuesday, March 14, 2006.

Probabilistic Risk Analysis (PRA) Evaluation ER14446, PRA Impact of [condition evaluationreport] 04-3262 Condition in the Significance Determine Process (SDP).QA-AUD-200414-0, Nonconformance Control and Corrective Actions Audit, dated 3/2/05.

Quality Assurance Surveillance QA-SUR-200438-0, Review of station policy concerninginsertion of manual reactor trips compared to the industry best practices.Reactor Trip Report - 03/30/04.

SA05-OD-05, Corrective Action Program Self-Assessment, Dec 5 - 9, 2005.

Self-Assessment Report; Assessment Number: SA04-OD-02, Corrective Action ProgramSelf-Assessment, October 11-14, 2004Significant Event Notice 257, Internal Flood Design Deficiencies.

Surveillance Test Task Sheet (STTS) Number 0410250, Diesel Generator A Operability Test.Surveillance Test Task Sheet No. 0415128 (for RHR pump B, completed October 14, 2004).Surveillance Test Task Sheet No. 0406219 (for Service Water (SW) Train A Valve OperabilityTest).Surveillance Test Task Sheet No. 0415128, RHR B Pump and Valve Operability Test(Group A).Surveillance Test Task Sheet No. 0521711, Component Cooling Pump A Comprehensive Test.

Surveillance Test Task Sheet No. 0510449, Control Room Train B Air Handling SystemOperation Test.Westinghouse Technical Bulletin TB-04-22, Revision 1: Reactor Coolant Pump SealPerformance and Appendix R Compliance.Westinghouse Infogram, IG-0-4, Nuclear Instrumentation Power Range Channel Detector Highvoltage Setting, April 28, 2004.Westinghouse Technical Bulletin TB-05-2, Potential Shorting of Printed Circuit Board Used inBarton Model 763A Pressure and Model 764 Differential Pressure Transmitters.Westinghouse Technical Bulletin TB-05-4, Potential Tin Whiskers on Printed Circuit BoardComponents.Work Request 9404338, Troubleshoot and correct (XVM02801B-MS, B Main Steam IsolationValve).Work Request 9404330, Repair the valve (XVM02801B-MS, B Main Steam Isolation Valve).

Work Order 0411249, Lube oil leak on the discharge of attached lube oil pump. (A DieselGenerator).Work Request 9404173, Adjust or repair the nuts on the guide rod plate.

(Nuclear Licensing) Evaluation of [condition evaluation report] 05-2300, Action 6, Feb., 2006.Technical Work Records38821, During STP-220.002\0317596, TDEFW pump test, rpm speed could not be adjustedbelow 4200 rpm.38828, This [condition evaluation report] generated for Maintenance Rule tracking purposes toprovide the 60-day completion of the functional failure evaluation from the failure of LD System Level Switch ILS01926 on PMTS 0313186.40201, Failure of DRCB/302 to close - [condition evaluation report] 04-1239, 4/27/04.CER 04-3767, Diesel-Driven Air Compressor Failure to Start.

A-13AttachmentCER 04-3775, Extraction Steam Pipe Failure.CER 05-1412, Diesel Driven Instrument Air Compressor Failed to Load.

CER 05-2286, Action 1 Evaluation.

CER 05-2622, Diesel Air Compressor Battery Failure Due to Loss of Level.

CR14992, EIR 81129 - Operating with standby condensate pump discharge valve open. Tab 2.14.ES 514, Cause Evaluation for IFV00498 Failing Closed.FM06742, XAA0029A(B) max resistance to preclude damper bypass, 11/8/05.

FM06742, Air flow instrumentation [condition evaluation report] 05-0975, 2/27/06.GR11674, CER 04-0015 Maintenance rule cause determination - turbine driven emergencyfeedpump governor valve linkage and speed control problem.JB17136, NCN (Non-conformance Notice) 99-0131, Disposition 18, Independent Review. Tab 00404.JB17136, NCN (Non-conformance Notice) 99-0131, Disposition 19, Independent Review. Tab 00405.LC13109, CER-04-1527, Action 6, Review of TB-04-22. Tab 615.

LS11312, CER/NCN 05-1891: Pipe Wall Thinning Associated With XVB03125B-SW. Tab 164.MC15002, [condition evaluation report] # 04-3786, Operability Evaluation. Tab MS.MT17700, ES-514 cause determination for the failures of ILS01902, ILS01905, and ILS01926, 5/27/04.MT17700, RCA 02-2500 for magnetrol level switches, 10/22/02.

MT17700, ES-514 Cause Determination for the failures of level switches 1LS01902, 1LS01929,and 1LS01966 ([condition evaluation report] 04-3230, Action #1, #2 & #3). Tab 3-32.MT17700, PZR (Pressurizer) Backup Group #2 Heater breaker XSW1DB 05 OperabilityRecommendation. Tab 3-33A.MT17700, ES-514 cause determination for the failures of ILT00459 on 4/11/04 ([conditionevaluation report] 04-1069, Action 2), 8/2/04.RM19910, NCN 03-3500: A D/G (Diesel Generator) (XEG0001A-E), Fuel Supply Pipe Broken atPipe Threads. File 01-49.WO 15209, NCN 05-3329, 50.59 Screening for accept-as-is. Tab EF/871.Work Orders0419265, Investigate cause of breaker failure.0510706, Charging pump B oil cooler component cooling water supply, packing leak.

0511749, Charging pump B oil cooler component cooling water supply, valve failed to close.

0511789, Charging pump B oil cooler component cooling water supply, breaker closing springnot charged.

A-14AttachmentLIST OF ACRONYMSADAMS-Agency Wide Documents Access and Management SystemCAP-Corrective Action Program CARB-Corrective Action Review Board CER-Condition Evaluation Report ECP-Employee Concerns Program ECR-Engineering Change Request INPO-Institute of Nuclear Power Operations IR-Inspection Report MPFFMaintenance Preventable Functional Failure MRT-Management Review Team MWR-Maintenance Work Request NCN-Nonconformance Notice NCV-Non-Cited Violation NL-Nuclear Licensing NRC-Nuclear Regulatory Commission OD&P-Organization Development & Performance OE-Operating Experience PIP-Primary Identification Program QA-Quality Assurance RCA-Root Cause Analysis SAP-Station Administrative Procedure SCE&G-South Carolina Electric & Gas Company SPB-Steam Propagation Barrier WO-Work Order