IR 05000270/2000287
| ML11333A342 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/28/2011 |
| From: | Bartley J H NRC/RGN-II/DRP/RPB1 |
| To: | Gillespie T P Duke Energy Carolinas |
| References | |
| Download: ML11333A342 (56) | |
Text
November 28, 2011
Mr. T. Preston Gillespie, Jr.
Site Vice President Duke Energy Carolinas, LLC Oconee Nuclear Station 7800 Rochester Highway Seneca, SC 29672
SUBJECT: PUBLIC MEETING SUMMARY - OCONEE NUCLEAR STATION - DOCKET NOS. 50-269, 50-270 AND 50-287
Dear Mr. Gillespie:
This refers to the Category 1 public meeting which was held on November 16, 2011, in Atlanta, GA. The purpose of this meeting was to discuss the safety significance of two preliminary greater than Green findings with two associated Apparent Violations (AV) that were documented in NRC Inspection Report 05000269,05000270, 05000287/2011018 (ML11277A253). A listing of meeting attendees and information presented during the meeting are enclosed.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRC'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).
Should you have any questions concerning this meeting, please contact me at (404) 997-4607.
Sincerely,/RA/ Jonathan H. Bartley, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos.: 50-269, 50-270, 50-287 License Nos.: DPR-38, DPR-47, DPR-55
Enclosures:
1. List of Attendees 2. NRC PowerPoint Presentation 3. Licensee PowerPoint Presentation cc w/encls: (See page 2)
November 28, 2011 Mr. T. Preston Gillespie, Jr.
Site Vice President Duke Energy Carolinas, LLC Oconee Nuclear Station 7800 Rochester Highway Seneca, SC 29672
SUBJECT: PUBLIC MEETING SUMMARY - OCONEE NUCLEAR STATION - DOCKET NOS. 50-269, 50-270 AND 50-287
Dear Mr. Gillespie:
This refers to the Category 1 public meeting which was held on November 16, 2011, in Atlanta, GA. The purpose of this meeting was to discuss the safety significance of two preliminary greater than Green findings with two associated Apparent Violations (AV) that were documented in NRC Inspection Report 05000269,05000270, 05000287/2011018 (ML11277A253). A listing of meeting attendees and information presented during the meeting are enclosed.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRC'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).
Should you have any questions concerning this meeting, please contact me at (404) 997-4607.
Sincerely,/RA/
Jonathan H. Bartley, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos.: 50-269, 50-270, 50-287 License Nos.: DPR-38, DPR-47, DPR-55
Enclosures:
1. List of Attendees 2. NRC PowerPoint Presentation 3. Licensee PowerPoint Presentation cc w/encls: (See page 2) X PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE X NON-SENSITIVE ADAMS: G Yes ACCESSION NUMBER:_ ML11333A342______________ G SUNSI REVIEW COMPLETE G FORM 665 ATTACHED OFFICE RII:DRP SIGNATURE /RA/ NAME JBartley DATE 11/28/2011 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: DOCUMENT4 DEC 2 cc w/encls: Division of Radiological Health TN Dept. of Environment & Conservation 401 Church Street Nashville, TN 37243-1532 Charles J. Thomas Fleet Licensing Manager Duke Energy Carolinas, LLC Electronic Mail Distribution David A. Baxter Vice President, Nuclear Engineering General Office Duke Energy Carolinas, LLC Electronic Mail Distribution David A. Cummings Associate General Counsel Duke Energy Corporation Electronic Mail Distribution Judy E. Smith Licensing Administrator Oconee Nuclear Station Duke Energy Carolinas, LLC Electronic Mail Distribution Kent Alter Regulatory Compliance Manager Oconee Nuclear Station Duke Energy Carolinas, LLC Electronic Mail Distribution Lara S. Nichols Vice President-Legal Duke Energy Corporation Electronic Mail Distribution Luellen B. Jones Fleet Licensing Engineer Duke Energy Carolinas, LLC Electronic Mail Distribution M. Christopher Nolan Fleet Safety Assurance Manager Duke Energy Carolinas, LLC Electronic Mail Distribution Sandra Threatt, Manager Nuclear Response and Emergency Environmental Surveillance Bureau of Land and Waste Management Department of Health and Environmental Control Electronic Mail Distribution Scott L. Batson Station Manager Oconee Nuclear Station Duke Energy Carolinas, LLC Electronic Mail Distribution Terry L. Patterson Safety Assurance Manager Duke Energy Carolinas, LLC Electronic Mail Distribution Charles Brinkman Director Washington Operations Westinghouse Electric Company, LLC Electronic Mail Distribution Tom D. Ray Engineering Manager Oconee Nuclear Station Duke Energy Carolinas, LLC Electronic Mail Distribution County Supervisor of Oconee County 415 S. Pine Street Walhalla, SC 29691-2145 W. Lee Cox, III Section Chief Radiation Protection Section N.C. Department of Environmental Commerce & Natural Resources Electronic Mail Distribution DEC 3 Letter to T. Preston Gillespie, Jr. from Jonathan H. Bartley dated November 28, 2011
SUBJECT: PUBLIC MEETING SUMMARY - OCONEE NUCLEAR STATION - DOCKET NOS. 50-269, 50-270 AND 50-287 Distribution w/encls: C. Evans, RII L. Douglas, RII OE Mail RIDSNRRDIRS PUBLIC RidsNrrPMOconee Resource Enclosure 1 2 Enclosure 1 Enclosure 2 OCONEE REGULATORY CONFERENCE November 16, 2011 2 Enclosure 2 Agenda*OPENING REMARKS AND INTRODUCTION*NRC REGULATORY AND ENFORCEMENT POLICY*STATEMENT OF ISSUES AND APPARENT VIOLATIONS*DUKE ENERGY CAROLINAS RESPONSE*BREAK/NRC CAUCUS*NRC FOLLOW UP QUESTIONS*CLOSING REMARKS*PUBLIC QUESTIONS 3 Enclosure 2 OCONEE REGULATORY CONFERENCE November 16, 2011 4 Enclosure 2 Issue #1A licensee-identified preliminary Greater Than Green Apparent Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified when the licensee failed to maintain design control of the Standby Shutdown Facility. The SSF pressurizerheater breakers and associated electrical components were not maintained as safety-related components nor seismically qualified as specified in the SSF licensing basis documents. Because the safety significance of this finding is preliminary Greater Than Green, it is being treated as an NRC-identified finding. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and adversely affected the cornerstone objective in that failure to maintain equipment qualification did not provide reasonable assurance that the Standby Shutdown Facility Auxiliary Service Water subsystem would perform its safety function. 10 CFR Part 50, Appendix B, Criteria III, Design Control, required, in part, that measures shall be established to assure that deviations from appropriate quality and design standards are controlled and that the review for suitability of application of equipment essential to safety-related functions of structures, systems, and components is maintained. Technical Specification 3.10.1.A required that, with the Standby Shutdown Facility Auxiliary Service Water inoperable, the system shall be restored to an operable status within seven days or the unit placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />.Contrary to the above, from 1983 until June 1, 2011, the licensee failed to review for suitability of application of equipment essential to safety-related functions of structures, systems, and components. The licensee failed to maintain the SSF pressurizerheater breakers and associated electrical components as safety-related QA-1 and seismically-qualified components in accordance with the licensing and design bases. As a result, the Standby Shutdown Facility was inoperable from 1983 until June 1, 2011, a period in excess of Technical Specification 3.10.1.A allowed outage time.The apparent violations discussed in this regulatory conference are subject to further review and are subject to change prior to any resulting enforcement action.
5 Enclosure 2 Issue #2An NRC-identified preliminary Greater Than Green Apparent Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensee's failure to install Standby Shutdown Facility pressurizer heater breakers that were qualified for expected environmental conditions inside of containment during design basis events. The licensee installed replacement breakers and the Standby Shutdown Facility was declared operable without testing to support that the replacement breakers would function under elevated containment temperatures. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control and adversely affected the cornerstone objective in that failure to maintain equipment qualification did not provide reasonable assurance that the Standby Shutdown Facility Auxiliary Service Water subsystem would perform its safety function. This finding had a cross-cutting aspect in the area of Human Performance under the Procedural Compliance aspect of the Work Practices component in that the licensee failed to follow the requirements set forth in EDM 601, Engineering Change H.4(b).10 CFR Part 50, Appendix B, Criteria III, Design Control, required, in part, that measures shall be established to assure that deviations from appropriate quality and design standards are controlled and that the review for suitability of application of equipment essential to safety-related functions of structures, systems, and components is maintained. Procedure EDM 601, Engineering Change, Section 601.5.2.1, Engineering Change Program -Design Phase, Commercial Controls, required that components acquired commercially be tested to verify they will perform the required safety function prior to actual installation. Technical Specification 3.10.1.A required that, with the Standby Shutdown Facility Auxiliary Service Water inoperable, the system shall be restored to an operable status within seven days or the unit placed in Mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. Contrary to the above, on June 4, 2011, a review for suitability of application of equipment essential to safety-related functions of structures, systems, and components was not performed. The licensee developed a modification that installed replacement breakers in the circuit supplying power to the pressurizer heaters from the Standby Shutdown Facility without test data to demonstrate that they would function at elevated containment temperatures and maintain the Standby Shutdown Facility functionality in accordance with the licensing and design bases. As a result, the Standby Shutdown Facility was inoperable from the time of new breaker installation (Unit 1 on June 8, 2011; Unit 2 on June 6, 2011; and Unit 3 on June 7, 2011) until August 20, 2011, a period in excess of Technical Specification 3.10.1.A allowed outage time.The apparent violations discussed in this regulatory conference are subject to further review and are subject to change prior to any resulting enforcement action.
6 Enclosure 2 Public Questions Enclosure 3 1Standby Shutdown Facility (SSF) Pressurizer Heater Breakers Units 1, 2, & 3NRC Region IIAtlanta, GeorgiaNovember 16, 2011Oconee Nuclear StationRegulatory Conference Enclosure 3 2Duke ParticipantsPreston GillespieOconee Site Vice PresidentTom RayOconee Engineering ManagerScott BatsonOconee Station ManagerTracy SavilleNGO Nuclear Engineering ManagerTerry PattersonOconee Safety Assurance ManagerBob GuyOconee Organizational Effectiveness ManagerBill PitesaSenior Vice President, Nuclear Operations Enclosure 3 3AgendaOpening RemarksP. GillespieOriginal Design ConfigurationT. RayTimeline, Breaker Selection & TestingT. RayOperations Role in the Operability DeterminationS. BatsonSignificance DeterminationT. SavilleRegulatory PerspectivesT. PattersonCause Analysis and Corrective ActionsB. GuyClosing RemarksB. Pitesa Enclosure 3 Opening Remarks4Preston Gillespie Oconee Site Vice President Enclosure 3 5Opening RemarksNRC proposed two findings related to the PZR HTRs powered from the SSF:A legacy issuewith the original design configurationA current performance issuerelated to the breakers installed and declared operableOconee agrees with NRC's characterization of the:Apparent violation of Appendix B, Criteria III (legacy issue)Apparent violation of Appendix B, Criteria III (current performance issue)Information will be presented to address:The self-identification of the SSF PZR HTR BKR problem in the original designThe immediate actions taken and actions taken to prevent recurrenceAssessment of the SSF design and licensing basesThe legacy issue alignment with the "Old Design Issue" criteria in IMC 0305 Enclosure 3 Opening RemarksNew or additional information is included to address the following:New test results for replacement breakers under a more representative environment Plant response to the loss of PZR HTRs in an SSF eventThermal-hydraulic modeling & confirming calculationsSimulator response consistent with analytical predictionsKey drivers contained in the probabilistic risk assessment:Human error probabilityBus duct fire analysisCable failure analysisPSV failure probability Oconee's determination of risk significance related to the findings are:Legacy issue -low to moderate safety significanceCurrent performance issue -very low safety significance6 Enclosure 3 Original Design Configuration7Tom RayOconee Engineer Manager Enclosure 3 SSF EquipmentASWInjects through OTSG upper nozzlesRC MakeupInjects through RCP seals29 gpm (nominal)SSF letdown linePressurizer heatersSubset of normal pressurizer heaters are powered from the SSF8RC MakeupASWSSF letdownHeaters Enclosure 3 9SSF PZR Heaters SchematicSimplified Schematic of SSF PZR Heater Circuitry Enclosure 3 SSF PZR Heaters Design Basis Oconee self-identified a deficiency with the SSF PZR HTR BKRsDeficiency identified during investigations associated with a future modificationContainment temperature increases on loss of building coolingThermal overloads are sensitive to increased ambient temperatureIndependent assessment of the SSF design & licensing bases The SSF is the system with the highest risk worth at OconeeThe assessment team includes independent and external membersContractors with previous SSFI and CDBI experience included on teamAssessment represents a sizable commitment of time and resourcesAssessment outcome will improve our performance10 Enclosure 3 11Timeline, Breaker Selection & TestingTom RayOconee Engineer Manager Enclosure 3 Timeline12(6/1) Concern Identified(6/2) SSF Inoperable Entered 7 Day LCO TS 3.10.1 Action A.1Installed GE Spectra Breakers June 6-8 for 3 UnitsSSF Restored via IDO & PDO for more testingSSF Declared Inoperable (3 Units)Entered 7 Day LCOTS 3.10.1 Action A.1Breakers replaced w/ qualified fusesSSF Declared OperableExpected completion of Spectra 225A and 70A Breakers under realistic profileV&V Complete for Water Solid Operations AOP Revision Jun 01 Jun 06Jun 08Jun 24Jul 08Jul 15Aug 20-21Oct 28Nov 153 of 4 Breakers trip at <267°F during LOCA TestingSSF Declared OBDNEntered TS 3.10.1 Action FSuccessful test of Spectra(70A) breaker using realistic profile Enclosure 3 Breaker SelectionRelative to BKR selection, Bounding Conditions for the SSF Function:The SSF mission time is 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sContainment temperature profile Loss of containment cooling with loss of all AC is the limiting conditionAging was addressed by limiting design lifeLow exposure levels based on location and event typeSelection of GE Spectra Breakers for Installation:Trip function less sensitive to temperatureBreaker function and materials were reviewed for suitabilityTesting was performed at a Duke facility to prove the BKRs would survive the high temperature ambient environmentBreakers were placed in service while testing activities continued13 Enclosure 3 Breaker Test ResultsTesting was conducted at Kinetics in June (GE Spectra Breaker):Testing was conducted in a LOCA chamber Breakers were powered during the testTesting profile increased temperature faster than actual conditionsTesting suspended when 3 of 4 failed Fuses were qualified at Kinectricsfor use as a permanent modificationRecent (November) testing is being conducted with the GE Spectra BreakersTesting with representative event temperature profileBreakers powered while in test oven at temperature (70A & 225A Breakers)Received report 11/14/11 from test lab the 225A Breaker trippedBreakers will be inspected and verified to trip before and after the 72 hr test14 Enclosure 3 SummaryEDM-601, Engineering Change Process, requires design inputs to be completed prior to declaring operableBased on preliminary insights from current root cause effort, weaknesses in implementing the modificationprocesses were as follows:Engineers had confidence in their engineering evaluation and testing was viewed as confirmatory rather than qualifyingPeer reviews did not challenge the implementation of the processWeaknesses in implementing the operability processesDecision-making influenced by previous use of water solid operationsModification improved but did not ensure the design function15 Enclosure 3 Operations Role in the Operability Determination16Scott BatsonOconee Station Manager Enclosure 3 17Standby Shutdown Facility Enclosure 3 Operability DeterminationsImmediate Determination of Operability -June 6 -8 , 2011OSM was knowledgeable of testing and resultsPeer review was obtained on the IDO prior to acceptance based upon reasonable assurance of operabilityTest verified temperature had no impact on trip settingsBreaker materials reviewed and supported the ambient insensitivity positionConfidence in Water Solid Operation as alternate means to meet pressure control functionRequired PDO to validate assumptions in the IDO18 Enclosure 3 Operability DeterminationsDetermination of Operability -June 24, 2011Licensed Operators trained that our analysis demonstrates that RCS pressure control from the SSF with Water Solid Operations is a technically sound and effective mitigation strategyOperating procedure guidance on Water Solid Operations since 2002Operating crews trained on Water Solid Operations in the SSF SimulatorProcedures were revised in 2011 to include additional identification stepsActions, indication and controls are simple and readily accessible in the SSF Control RoomSufficient time is available to complete operator actions19 Enclosure 3 InsightsWeakness in implementation of the Operability Determination ProcessDid not challenge process / procedure requirementsPeer review did not challenge process / procedure requirementsDecision-making influenced by previous use of Water Solid OperationsDid not recognize the need for prior NRC approval for Water Solid OperationFailed to recognize Water Solid Operation as a Compensatory MeasureActions were inconsistent with NSD 203, "Operability Determination Process"Inappropriately credited existing procedural guidanceA 10 CFR 50.59 review should have been performedOperability Determination Process improvement project will incorporate these insights.20 Enclosure 3 Significance Determination21Tracy SavilleNGO Nuclear Engineering Manager Enclosure 3 22Significance DeterminationOverviewSSF Operation for Normal/Loss of Heater ConditionsPRA Considerations for Water Solid Operations (WSO)Operating Procedural Flow PathsKey Differences with NRCHuman Reliability AnalysisBus Duct Fire AnalysisCable Failure AnalysisPSV failure probability Adjustments for Current Performance IssueRisk Results Enclosure 3 Initial SSF Activation & System ResponseSSF manned and activated, diesel generator startedASW provides heat sink to SGsPZR heaters energized to maintain RCS pressureRC makeup pump provides RCP seal cooling/RCS makeupASW flow controlled to RCS pressure setpointRCS slowly cools to offset RC makeupRCS trends to stable hot standby conditionT-cold ~550F (controlled by Main Steam relief valves)Natural circulation SG water levelsStable RCS pressure & PZR levelSSF letdown is cycled to control PZR level23 Enclosure 3 Response to Loss of HeatersPlant can be controlled via three procedural flow paths 1.Throttle ASW to control to RCS pressure (1950-2250 psig) with letdown line initially open2.Throttle ASW to maintain SG level setpoint and cycle letdown line block valve to maintain RCS pressure 1600-
2200 psig (most probable strategy used) 3.Hybrid, begins with path 1, transitions to path 2 once effect of open letdown line is recognized Strategy used is dependent upon conditions at time of heater loss and recognition of the symptom24 Enclosure 3 Operator Actions in WSOProcedural guidance and operator action times are sufficient to maintain RCS pressure within prescribed limits of 1600-2200 psig T/H calculations show: (Path 2 & Path 3)Pressure increase 1600 -2200 psig > 20 minPressure increase 2200 -2500 psig > 10 minPressure decrease 2200 -1600 psig > 90 minPlant simulator used to validate operator's ability to control the plant in WSO25 Enclosure 3 RCS Pressure Trace -WSO 26 Enclosure 3 PRA Considerations for WSOWSO is a viable means of maintaining the RCS in a safe and stable conditionBased on extensive Duke analysis which has been provided to NRCRisk significance of WSO following total heater loss is low and involves:Operator failure to control RCS pressure within prescribed procedural limitsComponent failure of letdown line block valve27 Enclosure 3 Procedural Flow Path Event Tree28SSF EventSSF SuccessfulHeaters SuccessfulProcedure PathComments0.000.201.000.700.10Path 1 - Throttle ASW to control to RCS pressure (1950-2250 psi) with letdown line initially openPath 2 - Throttle ASW to maintain SG level setpoint and cycle letdown line block valve to maintain RCS pressure 1600-2200 psiPath 3 - Begins like path 1 and transitions to path 2 once effect of open letdown line is recognizedBase Case - No delta riskBase Case - No delta risk Enclosure 3 Key DifferencesHuman Reliability Analysis (HRA)Detailed modeling of each procedural flow pathDuke uses an accepted industry analysis approach employed in the EPRI HRA Calculator ToolFor Path 2, the combined Human Error Probability is approximately 0.018.NRC used a screening value of 0.1A factor of 5.5 difference in resultsDuke applied a probability of 0.1 for the PSV failureThis is same probability the NRC used in previous ONS SDPNRC used a screening value of 1.029 Enclosure 3 Key DifferencesBus Duct Fire Initiating Event FrequencyDuke's analysis of record used a plant specific value of 1.2E-03 based on an EPRI data analysis reportThe NRC Phase III analysis used a generic frequency of 3.3E-03 from an NFPA-805 FAQThe NRC used a value of 2.4E-03 in a final significance determination in 2010 Duke believes the Oconee specific Bus Duct fire initiation probability the NRC used in its previous risk analysis is a better estimate than the generic NFPA-805 FAQ Bus Duct frequency used in the screeningOur results reflect the use of the plant specific value docketed in the NRC final significance determination in 2010 30 Enclosure 3 Key DifferencesCable Failure AnalysisDuke is performing a detailed cable failure analysis to more accurately determine the likelihood of a blackout.Duke has preliminary results which showed:A significant reduction in the Unit 3 Blackout FrequencyA modest reduction in the Unit 1 Blackout FrequencyInsignificant change in the Unit 2 Blackout FrequencyNRC analysis assumes that a station blackout would occur if these cable trays were affected.Preliminary information has been recently transmitted. Final report will be transmitted when finalized (~Nov. 18)31 Enclosure 3 Adjustments forCurrent Performance IssueReduced exposure period of ~75 days vs. 12 monthsEvent tree limited to Path 3 based on procedure changes made on June 6 Has lower failure probability than Path 2Fewer valve strokes and required operator actions Reflects lower tornado activity during exposure period (~June -August)Duke expects a lower breaker failure probability based on preliminary test data (sensitivity performed).32 Enclosure 3 Risk ResultsLegacy Issue (using NRC's 2010 bus duct frequency):Unit 1 -low E-06 Unit 2 -low E-06 Unit 3 -low E-06 Current Performance IssueMid E-07 (assuming 100% breaker failure probability)Mid E-08 (assuming 10% breaker failure probability)Risk associated with SSF heater failure is much lower than the NRC Phase III SDP analysis.33 Enclosure 3 Regulatory Perspectives34Terry PattersonOconee Safety Assurance Manager Enclosure 3 Regulatory Perspectives:Legacy IssueApparent Violation of Appendix B, Criterion III, Design Control:Oconee agrees with the apparent violationCorrective actions have been taken to restore complianceComprehensive actions taken or planned for extent-of-conditionSignificance:This issue has a low to moderate risk significanceOconee has offered its risk insights regarding the PRA analysis:Human Reliability AnalysisWater solid operations as a method of RCS pressure controlCable tray interaction with bus duct fires35 Enclosure 3 Old Design Issue Criteria[IMC 0305 Section 11.05(a)]Criterion 1 Licensee Identified:Oconee engineers identified the potential for similar conditions during modification activities in an unrelated area This behavior demonstrates a strong questioning attitudeThe condition was promptly entered into the corrective action programCriterion 2 -Immediate and long-term corrective actions to prevent recurrenceImmediate actions were prompt and comprehensiveCondition was addressed for all three unitsAction statement entered for all units within hours of identificationImprovements from the CAP Improvement project realizedQualified fuses installed to correct condition36 Enclosure 3 Criterion 3 -Opportunity for Prior Discovery:The degraded condition was not self-revealingThe degraded condition could not be observed during normal operations, routine testing or maintenance. Operating Experience review Criterion 4 -Current Performance:The breakers with thermal overloads had been installed since original constructionIdentification was the result of a strong questioning attitudeThe deficient condition was promptly entered into CAP and evaluated for operabilityImmediate actions have been taken and long term actions are in progress37Old Design Issue Criteria[IMC 0305 Section 11.05(a)]
Enclosure 3 Regulatory Perspectives -Current Performance IssueApparent Violation of Appendix B, Criterion III, Design Control:Oconee agrees with the apparent violationCorrective actions have been taken to restore complianceComprehensive actions taken or planned for extent-of-condition Significance:Reduced exposure period of ~75 days vs. 12 monthsProcedure changes implemented for WSOConducting test of Spectra BKRs with revised temperature profileThis issue has a very low risk significance38 Enclosure 3 Cause Analysis & Corrective Actions 39Bob GuyOrganizational Effectiveness Manager Enclosure 3 Legacy Issue: Cause DeterminationPreliminary Evaluation of Initial SSF Design Scoping 1978-1982Engineering review of the capabilities of re-purposed components (pressurizer heaters) to function under new conditions was not performed.Scoping of the project was inadequately specified and excluded the consideration of the pressurizer heater qualifications.Existing practice did not analyze re-purposed equipment to ensure that it would function under any new conditions.40 Enclosure 3 Legacy Issue:Corrective ActionsExtent of ConditionPressurizer heater panel board breakers -associated power circuitryEquipment breakers that must support safety equipment that will not function due to unexpected tripping due to environmental conditionsResolution of over 70 issuesCorrective ActionsDesign Change ProcessesDocumentation and TrainingOverall Independent Assessment (IA) of the SSF Two-phase approach -SSF Current Licensing Basis (CLB) Submittals and modifications41 Enclosure 3 Current Performance Issue:Cause AnalysisRoot Cause EvaluationCombined team to fully assess:Process applied during Engineering ChangesEngineering Changes for GE Spectra circuit breakersOperability Determination applicationOperability Determination process42 Enclosure 3 Current Performance IssueCorrective ActionsSafety Culture EnhancementsCAP Improvement ProjectImplemented Fleet wide -Industry benchmark information incorporatedStrengthened our questioning attitude and lowered entry thresholdsStrengthened our screening process, cause analysis tools, and trainingOperability Improvement ProjectFleet wide implementation -incorporate industry benchmark informationAlign IDO and PDO process with industry practiceProcedure Use & Adherence (PU&A)Engineering Change Manual Conservative Decision-Making43 Enclosure 3 Closing Remarks44Bill PitesaSenior Vice President, Nuclear Operations