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| | number = ML13205A214 | | | number = ML13205A214 |
| | issue date = 07/30/2013 | | | issue date = 07/30/2013 |
| | title = Wolf Creek Nuclear Operating Corporation - Slides for Pre-application Meeting on 7/30/13 - Core Design and Safety Analysis Methodology Transition License Amendment Request for Wolf Creek Generating Station (TAC No. MF2309) | | | title = Operating Corporation - Slides for Pre-application Meeting on 7/30/13 - Core Design and Safety Analysis Methodology Transition License Amendment Request for Wolf Creek Generating Station |
| | author name = | | | author name = |
| | author affiliation = Wolf Creek Nuclear Operating Corp | | | author affiliation = Wolf Creek Nuclear Operating Corp |
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| | docket = 05000482 | | | docket = 05000482 |
| | license number = NPF-042 | | | license number = NPF-042 |
| | contact person = Lyon C F | | | contact person = Lyon C |
| | case reference number = TAC MF2309 | | | case reference number = TAC MF2309 |
| | document type = Slides and Viewgraphs, Meeting Briefing Package/Handouts | | | document type = Slides and Viewgraphs, Meeting Briefing Package/Handouts |
| | page count = 29 | | | page count = 29 |
| | project = TAC:MF2309 | | | project = TAC:MF2309 |
| | stage = Other | | | stage = Meeting |
| }} | | }} |
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| =Text= | | =Text= |
| {{#Wiki_filter:WCNOC-NRCPre-submittalMeetingWCNOCNRC Presubmittal MeetingCore Design and Safety Analysis Methodology TransitionLicenseAmendmentRequestTransition License Amendment RequestJuly302013July 30, 2013Wolf Creek Nuclear Operating Corporationpgp1 Meeting AgendaMtiP/Objti*Meeting Purpose / Objectives*IntroductionsLicenseAmendmentRequest(LAR)Content*License Amendment Request (LAR) Content*Transition to Westinghouse Analysis MethodologiesMethodologies*RTS/ESFAS/LOP DG Start Instrumentation Setpoint Uncertainty Analysisyy*Alternative Source Term (AST)*LAR Schedule2*Summary2 Introductions*NRC*WCNOC Team *Westinghouse Support | | {{#Wiki_filter:WCNOC-NRC WCNOC NRC Pre-submittal Pre submittal Meeting Core Design and Safety Analysis Methodology Transition License Amendment Request July 30 30, 2013 Wolf Creek Nuclear Operating p g Corporation p |
| *Teleconference Attendees33 WCNOC Ownership of Safety AnalysisWCNOCOhifthSftAliidttd*WCNOC Ownership of the Safety Analysis is demonstrated through the strength and application of the Quality Assurance (QA) Program-WCNOC technical oversight is provided by the QA processes and participation in NUPIC-WCNOC QATittlfDiIftifdtth*Transmittal of Design Information -referenced to the source calculation, design basis document, or identified as a direct input*Owner's Acceptance Review | | 1 |
| *ConfigurationManagementupdates,regulatoryreviews,etc.Configuration Management updates, regulatory reviews, etc.*WCNOC/Westinghouse (WEC) Interface-On-site WEC staff responsible for core design and safety analysisOnsiteWECstaffmaintainsqualificationsinboththeWCNOCand44-On-site WEC staff maintains qualifications in both the WCNOC and WEC QA programs. Proper independence is always maintained between the QA programs.4 License Amendment Request Content*License Amendment Request (LAR) to revise the WCGS Technical Specifications (TSs) based on:-Transition to Westinghouse core design and safetyanalysismethodologiessafety analysis methodologies-Transition to Westinghouse Setpoint Methodology-FullScopeImplementationofAlternativeSourceFull Scope Implementation of Alternative Source Term (AST)55 License Amendment Request Content*LAR is consistent with NEI 06-02, "License Amendment Request (LAR) Guidelines" template Attachments:Attachments:I-EvaluationII-Proposed TS Changes (Mark-up) pg(p)III -Revised TS pages IV-Proposed TS Bases Changes (info only)V-Proposed COLR Changes (info only)6 License Amendment Request Content*LARcontent(cont)*LAR content (cont.) | | |
| | Meeting Agenda |
| | * Meeting M ti PPurpose / Objectives Obj ti |
| | * Introductions |
| | * License Amendment Request (LAR) Content |
| | * Transition to Westinghouse Analysis Methodologies |
| | * RTS/ESFAS/LOP DG Start Instrumentation Setpoint Uncertaintyy Analysis y |
| | * Alternative Source Term (AST) |
| | * LAR Schedule |
| | * Summary 2 |
| | |
| | ===Introductions=== |
| | * NRC |
| | * WCNOC Team |
| | * Westinghouse Support |
| | * Teleconference Attendees 3 |
| | |
| | WCNOC Ownership of Safety Analysis |
| | * WCNOC O Ownership hi off ththe S Safety f t AAnalysis l i iis d demonstrated t t d through the strength and application of the Quality Assurance (QA) Program |
| | - WCNOC technical oversight is provided by the QA processes and participation in NUPIC |
| | - WCNOC QA |
| | * T Transmittal itt l off Design D i Information I f ti - referenced f d tto the th source calculation, design basis document, or identified as a direct input |
| | * Owners Acceptance Review |
| | * Configuration Management updates, regulatory reviews, etc. |
| | * WCNOC/Westinghouse (WEC) Interface |
| | - On-site WEC staff responsible for core design and safety analysis |
| | - On-site On site WEC staff maintains qualifications in both the WCNOC and WEC QA programs. Proper independence is always maintained between the QA programs. 4 |
| | |
| | License Amendment Request Content |
| | * License Amendment Request (LAR) to revise the WCGS Technical Specifications (TSs) based on: |
| | - Transition to Westinghouse core design and safety analysis methodologies |
| | - Transition to Westinghouse Setpoint Methodology |
| | - Full Scope Implementation of Alternative Source Term (AST) 5 |
| | |
| | License Amendment Request Content |
| | * LAR is consistent with NEI 06-02, License Amendment Request (LAR) Guidelines template Attachments: |
| | I - Evaluation II - Proposed p TS Changes g ((Mark-up) p) |
| | III - Revised TS pages IV - Proposed TS Bases Changes (info only) |
| | V - Proposed COLR Changes (info only) 6 |
| | |
| | License Amendment Request Content |
| | * LAR content (cont.) |
| | (cont ) |
|
| |
|
| ==Enclosures:== | | ==Enclosures:== |
| I-WCAP-17658-NP-TransitionLicensingReportIWCAP17658NP Transition Licensing ReportII-WCAP-17746-P -Setpoint Methodology for WCGSIII -WCAP-17746-NPIV-WCAP-17602-P -Setpoint Calculations for WCGSV-WCAP-17602-NPVIFullScopeImplementationofAlternativeSourceTermVI-Full Scope Implementation of Alternative Source TermVII -CD-ROM containing Meteorological Data VIII -Proprietary Information Affidavit for WCAP-17746-PIX -Proprietary Information Affidavit for WCAP-17602-P77 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesCDidSftAliMthdl*Core Design and Safety Analysis Methodology-The Non-LOCA safety analyses were analyzed with Westinghouse, NRC approved methodsg,pp-All of the Westinghouse Non-LOCA methods are applicable to the WCGSTheContainmentresponseanalysesofrecordarenot-The Containment response analyses of record are not impacted-The SBLOCA and LBLOCA analyses of record are not impacted-The Core Design and Fuel Rod Design will be evaluated for each reload cycle88y8 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesOifAliMthdlS*Overview of Analysis Methodology Scope-Non-LOCA Safety Analyses*RETRAN-02 for Westinghouse PWRs (WCAP-14882-P-A) g()was used for majority of the analyses*LOFTRAN (WCAP-7907-P-A) was used for some analyses | | |
| *Other Codes that were used:-TWINKLE (WCAP-7979-P-A)-FACTRAN (WCAP-7908-A) -Non-LOCA Thermal-Hydraulics (T&H) Safety Analyses*VIPRE-01 (WCAP-14565-P-A) was used for the T&H analyses-RETRAN-02 (WCAP-14882-P-A) and WCAP-10698-P-A dfthStGtTbRtMi99were used for the Steam Generator Tube Rupture Margin to Overfill and Input to the Dose analyses9 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesOifAliMthdlS(t)*Overview of Analysis Methodology Scope (cont.)-DNB Correlations used in the VIPRE-01 DNBR calculationscalculations*The WRB-2 DNB correlation will continue to be used as the primary DNB correlation for the T&H analyses of fuel regions abovethefirstmixingvanegridabove the first mixing vane grid*The ABB-NV DNB correlation was used for the T&H analyses of fuel regions below the first mixing vane gridTheWLOPDNBcorrelationwasusedfortheT&Hanalysesthat*The WLOP DNB correlation was used for the T&H analyses that are outside the range of applicability of the WRB-2 and ABB-NV DNB correlations 101010 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesfOC*Implementation of the Westinghouse Non-LOCA Safety Analysis Methodology resulted in five changes tothecurrentWCGSTSsto the current WCGS TSs-SLs 2.1.1, Added the ABB-NV and WLOB DNB Correlations-TS 3.3.1, RTS Function 10, Reactor Coolant Flow-Low-TS 3.4.1, RCS Pressure, Temperature and Flow DNB LimitsLimits*The Minimum Measured Flow (MMF) was relocated to the COLR and revised from 371,000 gpm to 376,000 gpm1111*The Thermal Design Flow -361,200 gpm, replaces the MMF11 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesIlttifthWtihNLOCASft*Implementation of the Westinghouse Non-LOCA Safety Analysis Methodology results in five changes to the current WCGS TSs (cont.)()-TS Table 3.7.1-1, OPERABLE MSSVs versus Maximum Allowable Power, the maximum allowable power for 4, 3, and2OPERABLEMSSVswasrevisedand 2 OPERABLE MSSVs was revised-TS 5.6.5, COLR*Added WCAP-9272-P-A, the Westinghouse Reload MthdltSifiti565bMethodology to Specification 5.6.5 b.*Deleted the WCNOC methodologies from Specification 5.6.5 b.*WCAP-9272-P-A is the only methodology associated with a Tech Spec COLR parameter1212 TransitiontoWestinghouseAnalysisTransition to Westinghouse Analysis MethodologiesFllfth9/20/12PSbittlMti*Followup from the 9/20/12 Pre-Submittal Meeting-Provide a roadmap of which code was used to analyze each postulated accident -Attachment I of the LAR includes a roadmap-The Limitations, Restrictions and Conditions for the Westinghouse codes used in the Non-LOCA safety analyses are addressed in detail, including justifications in the LAR (Enclosure I, Appendix A)131313 InstrumentationSetpointUncertaintyInstrumentation Setpoint Uncertainty AnalysisUtitAli*Uncertainty Analysis-Transition from the existing WCNOC Setpoint Methodology to the current Westinghouse Setpoint gygpMethodology as applied to WCGS for RTS, ESFAS and LOP DG Start Instrumentation (WCAP-17746-P, Enclosure II of LAR),)-Technical Specification Changes*TS 3.3.1,TS 3.3.2, and TS 3.3.5 Allowable Values were replacedwithaNominalTripSetpointreplaced with a Nominal Trip Setpoint*TS Table 3.3.1-1, Overtemperature T, Note 1 and Overpower T, Note 2 14 1414 InstrumentationSetpointUncertaintyUtitAli(t)Instrumentation Setpoint Uncertainty Analysis*Uncertainty Analysis (cont.)-Calculations were performed for the RTS, ESFAS, and LOP DG Start instrumentation Functions using the current Westinghouse setpoint methodology (WCAP-17602-P, Enclosure IV of LAR))-Implementation of the Westinghouse Setpoint Methodology resulted in two changes to the existingWCGSTripSetpointsexisting WCGS Trip Setpoints*TS 3.3.1, RTS Function 10, Reactor Coolant Flow-Low*TS 3.3.5, Degraded Voltage Function1515 InstrumentationSetpointUncertaintyInstrumentation Setpoint Uncertainty AnalysisTSTF493ARii4OtiAThil*TSTF-493-A, Revision 4, Option A -Technical Specification Changes-VariationfromOptionA-NominalTripSetpointVariation from Option A Nominal Trip Setpoint specified in the single column format based on the Westinghouse Setpoint MethodologyTShildthdditifidiidl-TS changes include the addition of individual Surveillance Requirement footnotes to the applicable instrumentation Functions in accordance with Option pA of TSTF-493, Revision 41616 InstrumentationSetpointUncertaintyInstrumentation Setpoint Uncertainty Analysis*Followup from the 9/20/12 Pre-Submittal Meeting-The level of detail of the setpoint methodology and setpoint calculations for WCGS is consistent with that in the Diablo Canyon Power Plant (DCPP) y()submittal of March 7, 2013 (DCL-13-016)-WCNOC specific setpoint calculations are providedinWCAP-17602-P(EnclosureIVofLAR)provided in WCAP17602P (Enclosure IV of LAR)17 Alternative Source TermFllSIlttifthAST*Full Scope Implementation of the AST-Radiological dose consequences analyses were performed for the accidents specified in Regulatory ppgyGuide (RG) 1.183 include:*Main Steamline Break (USAR Section 15.1.5.3)*LockedRotor(USARSection15333)*Locked Rotor (USAR Section 15.3.3.3)*Rod Ejection (USAR Section 15.4.8.3)*Steam Generator Tube Rupture (USAR Section 15.6.3.3)*Loss of Coolant Accident (USAR Section 15.6.5.4)*Fuel Handling Accident (USAR Section 15.7.4)181818 Alternative Source Term*Full Scope Implementation of the AST (cont.)-Radiological dose consequences analyses performedforadditionalaccidentsnotspecifiedinperformed for additional accidents not specified in RG 1.183 include:*Loss of Non-Emergency AC Power (USAR Section 15263)15.2.6.3)*Letdown Line Break (USAR Section 15.6.2.1)*Waste Gas Decay Tank Failure (USAR Section 15.7.1)LiqidWasteTankFailre(USARSection1572)*Liquid Waste Tank Failure (USAR Section 15.7.2)-Dose consequences analyses were performed using version 3.03 of the RADTRAD computer dcode1919 Alternative Source Term*Full Scope Implementation of AST (cont.)-No changes to the licensing basis EQ dose analysesmaintainingtheTID14844accidentanalyses -maintaining the TID-14844 accident source term -No changes to the licensing basis NUREG-0737 evaluations other than the Control Room Habitability Envelope (CHRE) doses (III.D.3.4) and Technical SupportCenterdoses(IIIA12)Support Center doses (III.A.1.2)202020 Alternative Source Term*Atmospheric Dispersion Factors (X/Q)-New X/Q values were calculated*Offsite(EABandLPZ)X/Qvalueswerecalculatedusing*Offsite (EAB and LPZ) X/Q values were calculated using the PAVAN code consistent with RG 1.145*Control Room and TSC X/Q values were calculated using theARCON96codeconsistentwithRG1194the ARCON96 code consistent with RG 1.194-Meteorological Data*Five years of WCGS site-specific meteorological data from 1/1/2006 through 12/31/2010 was collected*Data recovery for the 5-year period met the 90% recovery criterion of RG 1.23212121 Alternative Source Term*Current licensing basis changes-Revises USAR Chapter 15 dose analysis for 10 accidents (includes the 6 DBAs in RG 1.183)()-New Offsite, Control Room, and TSC atmospheric dispersion factors based on site-specific meteorological data from 2006 through 2010g-Revises the CRHE unfiltered inleakage from 20 scfm to 50 scfm-Revises the Control Building unfiltered inleakage from 300 scfm to 400scfm400 scfm-TS changes to address the update of the accident source term and associated DBAsTSchangestoaddresstheadoptionofTSTF51ARevision22222-TS changes to address the adoption of TSTF-51-A, Revision 222 Alternative Source TermThilSifitiCh*Technical Specification Changes-Definition of DOSE EQUIVALENT I-131*RevisedtoonlyallowtheuseofthedoseconversionfactorsRevised to only allow the use of the dose conversion factors from EPA Federal Guidance Report No. 11 -Definition of DOSE EQUIVALENT XE-133*RevisedtoonlyallowtheuseofthedoseconversionfactorsRevised to only allow the use of the dose conversion factors EPA Federal Guidance Report No. 12 -Specification 5.5.12, "Explosive Gas and Storage TankRadioactivityMonitoringProgram"Tank Radioactivity Monitoring Program*Revises the quantity of radioactivity contained in each gas storage tank to be less than the amount that would result in a whole body exposure limit to 0.1 rem (current limit is 0.5 23 23yp(rem) 23 Alternative Source TermThilSifitiCh(t)*Technical Specification Changes (cont.)-Adoption of TSTF-51-A, Revision 2, "Revise Containment Requirements during Handling qggIrradiated Fuel and Core Alterations"*Allows the elimination of the TS requirements for certain Engineered Safety Feature (ESF) systems to be gy()yOPERABLE, after a sufficient radioactive decay has occurred*Changes were not applied to the TS Section 3.8 Electrical TSs (conservative)242424 Alternative Source TermElVIfLARSti*Enclosure VI of LAR -Sections1 -DESCRIPTION2 -PROPOSED CHANGES3 -BACKGROUND 4 -TECHNICAL ANALYSIS 5 -RG 1.183 CONFORMANCE TABLE6 -RG 1.145 CONFORMANCE TABLE 7 -RG 1.194 CONFORMANCE TABLE 8 -RIS 2006-04 TABLE9 -PROPOSED TS MARKUPS 10 -RETYPED TS PAGES11-PROPOSEDBASESMARKUPS(informationonly)252511 PROPOSED BASES MARKUPS (information only)12 -PROPOSED TRM and BASES MARKUP (information only)13 -PROPOSED USAR CHANGES (information only)25 Alternative Source Term*Followup from the 9/20/12 Pre-Submittal Meeting-Meteorological Data -One gap in the recorded data was due to the data logger failure (5/30/2007 through6/7/2007)through 6/7/2007)-Provide a detailed plant drawing that shows the potential release paths -site plan provided consistent with the guidance in RIS 2006-04 (Enclosure VI of the LAR)26 Alternative Source Term*Followupfrom the 9/20/12 Pre-Submittal Meeting(cont)27 Schedule*Submit LAR to NRCAugust 13, 2013*NRC Acceptance Review (30 days per LIC-109)*Requested Approval DateDecember 15, 2014*Start of Refueling Outage 20January 5, 2015*Cycle 21 StartupFebruary 9, 20152828 SummaryWCNOCitdtbitLAR8/13/13ti*WCNOC intends to submit a LAR on 8/13/13 to revise the WCGS TSs based on:-TransitiontotheWestinghousecoredesignandTransition to the Westinghouse core design and safety analysis methodologies-Transition to the Westinghouse Setpoint Methodology-Full Scope Implementation of the Alternative SourceTerm(AST)Source Term (AST)*Request NRC approval by 12/15/14 to support Cycle 21 operation (Feb. 2015)29*Questions/Comments29 | | I - WCAP 17658 NP - Transition Licensing Report WCAP-17658-NP II - WCAP-17746-P - Setpoint Methodology for WCGS III - WCAP-17746-NP IV - WCAP-17602-P - Setpoint Calculations for WCGS V - WCAP-17602-NP VI - Full Scope Implementation of Alternative Source Term VII - CD-ROM containing Meteorological Data VIII - Proprietary Information Affidavit for WCAP-17746-P IX - Proprietary Information Affidavit for WCAP-17602-P 7 |
| }} | | |
| | Transition to Westinghouse Analysis Methodologies |
| | * Core C D Designi and dS Safety f t Analysis A l i Methodology M th d l |
| | - The Non-LOCA safety analyses were analyzed with g |
| | Westinghouse, , NRC approved pp methods |
| | - All of the Westinghouse Non-LOCA methods are applicable to the WCGS |
| | - The Containment response analyses of record are not impacted |
| | - The SBLOCA and LBLOCA analyses of record are not impacted |
| | - The Core Design and Fuel Rod Design will be y |
| | evaluated for each reload cycle 8 |
| | |
| | Transition to Westinghouse Analysis Methodologies |
| | * Overview O i off Analysis A l i M Methodology th d l S Scope |
| | - Non-LOCA Safety Analyses |
| | * RETRAN-02 for Westinghouse g PWRs ((WCAP-14882-P-A)) |
| | was used for majority of the analyses |
| | * LOFTRAN (WCAP-7907-P-A) was used for some analyses |
| | * Other Codes that were used: |
| | - TWINKLE (WCAP-7979-P-A) |
| | - FACTRAN (WCAP-7908-A) |
| | - Non-LOCA Thermal-Hydraulics (T&H) Safety Analyses |
| | * VIPRE-01 (WCAP-14565-P-A) was used for the T&H analyses |
| | - RETRAN-02 (WCAP-14882-P-A) and WCAP-10698-P-A were used d ffor th the Steam St Generator G t Tube T b Rupture R t Margin M i to Overfill and Input to the Dose analyses 9 |
| | |
| | Transition to Westinghouse Analysis Methodologies |
| | * Overview O i off Analysis A l i M Methodology th d l S Scope ((cont.) |
| | t) |
| | - DNB Correlations used in the VIPRE-01 DNBR calculations |
| | * The WRB-2 DNB correlation will continue to be used as the primary DNB correlation for the T&H analyses of fuel regions above the first mixing vane grid |
| | * The ABB-NV DNB correlation was used for the T&H analyses of fuel regions below the first mixing vane grid |
| | * The WLOP DNB correlation was used for the T&H analyses that are outside the range of applicability of the WRB-2 and ABB-NV DNB correlations 10 |
| | |
| | Transition to Westinghouse Analysis Methodologies |
| | * Implementation off the Westinghouse Non-LOCA OC Safety Analysis Methodology resulted in five changes to the current WCGS TSs |
| | - SLs 2.1.1, Added the ABB-NV and WLOB DNB Correlations |
| | - TS 3.3.1, RTS Function 10, Reactor Coolant Flow- Low |
| | - TS 3.4.1, RCS Pressure, Temperature and Flow DNB Limits |
| | * The Minimum Measured Flow (MMF) was relocated to the COLR and revised from 371,000 gpm to 376,000 gpm |
| | * The Thermal Design Flow - 361,200 gpm, replaces the MMF 11 |
| | |
| | Transition to Westinghouse Analysis Methodologies |
| | * IImplementation l t ti off theth Westinghouse W ti h Non-LOCA N LOCA Safety S f t Analysis Methodology results in five changes to the current WCGS TSs (cont.) ( ) |
| | - TS Table 3.7.1-1, OPERABLE MSSVs versus Maximum Allowable Power, the maximum allowable power for 4, 3, and 2 OPERABLE MSSVs was revised |
| | - TS 5.6.5, COLR |
| | * Added WCAP-9272-P-A, the Westinghouse Reload M th d l Methodology tto S Specification ifi ti 5 5.6.5 65b b. |
| | * Deleted the WCNOC methodologies from Specification 5.6.5 b. |
| | * WCAP-9272-P-A is the only methodology associated with a Tech Spec COLR parameter 12 |
| | |
| | Transition to Westinghouse Analysis Methodologies |
| | * Followup F ll ffrom th the 9/20/12 P Pre-Submittal S b itt l M Meeting ti |
| | - Provide a roadmap of which code was used to analyze each postulated accident - Attachment I of the LAR includes a roadmap |
| | - The Limitations, Restrictions and Conditions for the Westinghouse codes used in the Non-LOCA safety analyses are addressed in detail, including justifications in the LAR (Enclosure I, Appendix A) 13 |
| | |
| | Instrumentation Setpoint Uncertainty Analysis |
| | * Uncertainty U t i t Analysis A l i |
| | - Transition from the existing WCNOC Setpoint Methodology gy to the current Westinghouse g Setpoint p |
| | Methodology as applied to WCGS for RTS, ESFAS and LOP DG Start Instrumentation (WCAP-17746-P,, Enclosure II of LAR)) |
| | - Technical Specification Changes |
| | * TS 3.3.1,TS 3.3.2, and TS 3.3.5 Allowable Values were replaced with a Nominal Trip Setpoint |
| | * TS Table 3.3.1-1, Overtemperature T, Note 1 and Overpower T, Note 2 14 |
| | |
| | Instrumentation Setpoint Uncertainty Analysis |
| | * Uncertainty U t i t Analysis A l i ((cont.) t) |
| | - Calculations were performed for the RTS, ESFAS, and LOP DG Start instrumentation Functions using the current Westinghouse setpoint methodology (WCAP-17602-P, Enclosure IV of LAR)) |
| | - Implementation of the Westinghouse Setpoint Methodology resulted in two changes to the existing WCGS Trip Setpoints |
| | * TS 3.3.1, RTS Function 10, Reactor Coolant Flow- Low |
| | * TS 3.3.5, Degraded Voltage Function 15 |
| | |
| | Instrumentation Setpoint Uncertainty Analysis |
| | * TSTF TSTF-493-A, 493 A Revision R i i 4, 4 Option O ti A - Technical T h i l Specification Changes |
| | - Variation from Option A - Nominal Trip Setpoint specified in the single column format based on the Westinghouse Setpoint Methodology |
| | - TS changes h include i l d the th addition dditi off individual i di id l Surveillance Requirement footnotes to the applicable instrumentation Functions in accordance with Option p |
| | A of TSTF-493, Revision 4 16 |
| | |
| | Instrumentation Setpoint Uncertainty Analysis |
| | * Followup from the 9/20/12 Pre-Submittal Meeting |
| | - The level of detail of the setpoint methodology and setpoint calculations for WCGS is consistent with that in the Diablo Canyon y Power Plant ((DCPP)) |
| | submittal of March 7, 2013 (DCL-13-016) |
| | - WCNOC specific setpoint calculations are provided in WCAP WCAP-17602-P 17602 P (Enclosure IV of LAR) 17 |
| | |
| | Alternative Source Term |
| | * Full F ll S Scope IImplementation l t ti off the th AST |
| | - Radiological dose consequences analyses were performed for the accidents specified p p in Regulatory g y Guide (RG) 1.183 include: |
| | * Main Steamline Break (USAR Section 15.1.5.3) |
| | * Locked Rotor (USAR Section 15.3.3.3) 15 3 3 3) |
| | * Rod Ejection (USAR Section 15.4.8.3) |
| | * Steam Generator Tube Rupture (USAR Section 15.6.3.3) |
| | * Loss of Coolant Accident (USAR Section 15.6.5.4) |
| | * Fuel Handling Accident (USAR Section 15.7.4) 18 |
| | |
| | Alternative Source Term |
| | * Full Scope Implementation of the AST (cont.) |
| | - Radiological dose consequences analyses performed for additional accidents not specified in RG 1.183 include: |
| | * Loss of Non-Emergency AC Power (USAR Section 15 2 6 3) 15.2.6.3) |
| | * Letdown Line Break (USAR Section 15.6.2.1) |
| | * Waste Gas Decay Tank Failure (USAR Section 15.7.1) |
| | * Liquid Liq id Waste Tank Fail Failure re (USAR Section 15.7.2) 15 7 2) |
| | - Dose consequences analyses were performed using version 3.03 of the RADTRAD computer code d |
| | 19 |
| | |
| | Alternative Source Term |
| | * Full Scope Implementation of AST (cont.) |
| | - No changes to the licensing basis EQ dose analyses - maintaining the TID TID-14844 14844 accident source term |
| | - No changes to the licensing basis NUREG-0737 evaluations other than the Control Room Habitability Envelope (CHRE) doses (III.D.3.4) and Technical Support Center doses (III (III.A.1.2) |
| | A 1 2) 20 |
| | |
| | Alternative Source Term |
| | * Atmospheric Dispersion Factors (X/Q) |
| | - New X/Q values were calculated |
| | * Offsite (EAB and LPZ) X/Q values were calculated using the PAVAN code consistent with RG 1.145 |
| | * Control Room and TSC X/Q values were calculated using the ARCON96 code consistent with RG 1 1.194 194 |
| | - Meteorological Data |
| | * Five years of WCGS site-specific meteorological data from 1/1/2006 through 12/31/2010 was collected |
| | * Data recovery for the 5-year period met the 90% recovery criterion of RG 1.23 21 |
| | |
| | Alternative Source Term |
| | * Current licensing basis changes |
| | - Revises USAR Chapter 15 dose analysis for 10 accidents |
| | ((includes the 6 DBAs in RG 1.183)) |
| | - New Offsite, Control Room, and TSC atmospheric dispersion factors based on site-specific meteorological data from 2006 throughg 2010 |
| | - Revises the CRHE unfiltered inleakage from 20 scfm to 50 scfm |
| | - Revises the Control Building unfiltered inleakage from 300 scfm to 400 scfm |
| | - TS changes to address the update of the accident source term and associated DBAs |
| | - TS changes to address the adoption of TSTF TSTF-51-A, 51 A Revision 2 22 |
| | |
| | Alternative Source Term |
| | * Technical T h i lS Specification ifi ti Ch Changes |
| | - Definition of DOSE EQUIVALENT I-131 |
| | * Revised to only allow the use of the dose conversion factors from EPA Federal Guidance Report No. 11 |
| | - Definition of DOSE EQUIVALENT XE-133 |
| | * Revised to only allow the use of the dose conversion factors EPA Federal Guidance Report No. 12 |
| | - Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring ProgramProgram |
| | * Revises the quantity of radioactivity contained in each gas storage tank to be less than the amount that would result in a whole bodyy exposure p limit to 0.1 rem ((current limit is 0.5 rem) 23 |
| | |
| | Alternative Source Term |
| | * Technical T h i lS Specification ifi ti Ch Changes ((cont.) t) |
| | - Adoption of TSTF-51-A, Revision 2, Revise q |
| | Containment Requirements during g Handling g Irradiated Fuel and Core Alterations |
| | * Allows the elimination of the TS requirements for certain g |
| | Engineered Safetyy Feature ((ESF)) systems y to be OPERABLE, after a sufficient radioactive decay has occurred |
| | * Changes were not applied to the TS Section 3.8 Electrical TSs (conservative) 24 |
| | |
| | Alternative Source Term |
| | * E l Enclosure VI off LAR - Sections S ti 1 - DESCRIPTION 2 - PROPOSED CHANGES 3 - BACKGROUND 4 - TECHNICAL ANALYSIS 5 - RG 1.183 CONFORMANCE TABLE 6 - RG 1.145 CONFORMANCE TABLE 7 - RG 1.194 CONFORMANCE TABLE 8 - RIS 2006-04 TABLE 9 - PROPOSED TS MARKUPS 10 - RETYPED TS PAGES 11 - PROPOSED BASES MARKUPS (information only) 12 - PROPOSED TRM and BASES MARKUP (information only) 25 13 - PROPOSED USAR CHANGES (information only) |
| | |
| | Alternative Source Term |
| | * Followup from the 9/20/12 Pre-Submittal Meeting |
| | - Meteorological Data - One gap in the recorded data was due to the data logger failure (5/30/2007 through 6/7/2007) |
| | - Provide a detailed plant drawing that shows the potential release paths - site plan provided consistent with the guidance in RIS 2006-04 (Enclosure VI of the LAR) 26 |
| | |
| | Alternative Source Term |
| | * Followup from the 9/20/12 Pre-Submittal Meeting (cont) 27 |
| | |
| | Schedule |
| | * Submit LAR to NRC August 13, 2013 |
| | * NRC Acceptance Review (30 days per LIC-109) |
| | * Requested Approval Date December 15, 2014 |
| | * Start of Refueling Outage 20 January 5, 2015 |
| | * Cycle 21 Startup February 9, 2015 28 |
| | |
| | Summary |
| | * WCNOC iintends t d tto submit b it a LAR on 8/13/13 tto revise i |
| | the WCGS TSs based on: |
| | - Transition to the Westinghouse core design and safety analysis methodologies |
| | - Transition to the Westinghouse Setpoint Methodology |
| | - Full Scope Implementation of the Alternative Source Term (AST) |
| | * Request NRC approval by 12/15/14 to support Cycle 21 operation (Feb. 2015) |
| | * Questions/Comments 29}} |
Letter Sequence Meeting |
---|
|
|
MONTHYEARML13179A1012013-07-0909 July 2013 7/30/13 Notice of Forthcoming Preapplication Meeting with Wolf Creek Nuclear Operating Company to Discuss Proposed Transition to Westinghouse Methodologies for Core Design and Non-LOCA Safety Analyses at Wolf Creek Generating Station Project stage: Meeting ML13205A2142013-07-30030 July 2013 Operating Corporation - Slides for Pre-application Meeting on 7/30/13 - Core Design and Safety Analysis Methodology Transition License Amendment Request for Wolf Creek Generating Station Project stage: Meeting ML13221A2152013-08-12012 August 2013 7/30/2013 - Summary of Preapplication Meeting with Wolf Creek Nuclear Operating Company to Discuss Proposed Transition to Westinghouse Methodologies for Core Design and Non-LOCA Safety Analyses at Wolf Creek Generating Station Project stage: Meeting 2013-07-09
[Table View] |
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Category:Slides and Viewgraphs
MONTHYEARML23275A1712023-10-16016 October 2023 October 16, 2023, Licensee Pre-submittal Meeting Slides - License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505-A, Revision 2 for Wolf Creek Generating Station ML21333A1512021-11-29029 November 2021 Pre-Submittal Meeting Slides License Amendment Request Addressing Portable Lighting for Operator Manual Actions ML20310A2232020-11-19019 November 2020 Slides for ERO Staffing pre-submittal Public Teleconference November 19, 2020 (L-2020-LRM-0104) IR 05000482/20174072017-12-22022 December 2017 Summary of Closed Regulatory Conference to Discuss Wolf Creek, Unit 1, Security Inspection Report 05000482/2017407 ML16236A0952016-08-25025 August 2016 and Wolf Creek Generating Station - August 25, 2016, Class 1E Electrical Equipment Air Conditioning System Pre-Application Meeting ML16236A0972016-08-25025 August 2016 Operating Corporation - August 25, 2016, Core Design and Safety System Analysis Methodology Transition License Amendment Request Revised ML16095A0782016-04-12012 April 2016 Slides for the WCGS Meeting Discussion 4/12/16 ML14188C4812014-07-0707 July 2014 Summary of Annual Assessment Meeting with Wolf Creek Generating Station ML14134A1842014-05-14014 May 2014 Summary of Public Meeting to Discuss an Apparent Violation Identified at Wolf Creek Generating Station ML14038A3852014-02-0606 February 2014 1/22/2014 - Summary of Public Meeting to Discuss Corrective Actions Implemented to Address the Chilling Effect Letter ML13218A0212013-08-0707 August 2013 Licensee Slides, 08/07/13 Preapplication Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Long-Term Corrective Actions for Water Hammer Events in the Essential Service Water System at Wolf Creek ML13219A1932013-08-0707 August 2013 Licensee Slides (Final), 08/07/13 Preapplication Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Long-Term Corrective Actions for Water Hammer Events in the Essential Service Water System at Wolf Creek ML13205A2142013-07-30030 July 2013 Operating Corporation - Slides for Pre-application Meeting on 7/30/13 - Core Design and Safety Analysis Methodology Transition License Amendment Request for Wolf Creek Generating Station ML13121A4892013-05-0101 May 2013 End of Cycle Meeting Summary 4-18-13 ML12263A3622012-09-20020 September 2012 Pre-application Meeting Slide Core Design and Safety Analysis Methodology Transition License Amendment Request TAC No. ME9495) ML12191A1742012-07-0909 July 2012 6/25/2012 Summary of Public Meeting with Wolf Creek Nuclear Operating Corporation ML1110104292011-04-11011 April 2011 Summary of Annual Performance Assessment Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Wolf Creek Generating Station Performance for the NRC Inspection Period Ending December 31, 2010 ML1016901162010-06-16016 June 2010 6/16/2010 Construction Reactor Oversight Process Category 2 Public Meeting Handout: Meeting Slides (Enforcement Cases) ML1013404892010-05-14014 May 2010 Summary of Annual Performance Assessment Meeting with Wolf Creek Nuclear Operating Corporation to Discuss the Wolf Creek Generating Station Performance for the NRC Inspection Period from January 1 Through December 31, 2009 ML0933702032009-12-0303 December 2009 Summary of Public Meeting for Wolf Creek Generating Station ML0933801482009-11-20020 November 2009 Licensee Handouts from November 20, 2009, Public Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss GL 2004-02 Response Rai'S ML0913300232009-05-13013 May 2009 Meeting Slides, Successful Licensing of the Als Fpga Based Safety Related I&C Platform ML0912600762009-05-0606 May 2009 Lessons Learned Using Digital I&C Interim Staff Guidance Workshop Application of ISG-4 During Wolf Creek and Oconee Reviews ML0912600692009-05-0606 May 2009 Lessons Learned Using Digital I&C Interim Staff Guidance Workshop ML0902102422009-01-15015 January 2009 Licensee Slides, January 15, 2009, Category 1 Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Changes to Post Fire Shutdown Unresolved Items Analysis Methods Pre-Application for Wolf Creek Generating Station ML0834700222008-12-11011 December 2008 Slides, Category 2 Public Meeting Digital Instrumentation and Control Steering Committee M080717, M080717-Commission Briefing Slides/Exhibits Briefing on Fire Protection2008-07-17017 July 2008 M080717-Commission Briefing Slides/Exhibits Briefing on Fire Protection ML0817706682008-06-25025 June 2008 Summary of Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Several Initiatives That Were Being Implemented to Improve Plant Performance at Wolf Creek ML0813606472008-06-10010 June 2008 05/01/2008-Summary of Public Meeting on Safety Evaluation Report with Open Items Regarding the Wolf Creek Generating Station License Renewal Review ML0721102382007-07-24024 July 2007 Performanc Contracting, Inc.'S Powerpoint Slide, Proposed Wolf Creek/Callaway Test Configuration ML0720601382007-07-17017 July 2007 Slides from Meeting Between NRC Staff and Wolf Creek Panel Enclosures 2 to 3, Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential Pwscc. ML0719303772007-07-11011 July 2007 Handouts (NRC and Licensee) for Meeting with Representatives of Wolf Creek Nuclear Operating Corporation ML0718606902007-06-19019 June 2007 06/19-20/2007 Slides,Fabrication Records Review, from Category 2 Public Meeting Between the NRC Staff and the Expert Panel for the Wolf Creek Advanced Finite Element Analyses (Fea) ML0716203882007-06-0101 June 2007 06/01/07 - Presentation Material, Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential Pwscc. ML0733409182007-05-31031 May 2007 05/31/2007 Presentation by J. Cudsworth NRC Treatment of Issues Other than Category 2 Issues for License Renewal ML0713703642007-05-17017 May 2007 Handout for Meeting with Wolf Creek Nuclear Operating Corporation on the Licensee'S Application for the Main Steam and Feedwater Isolation System (Msfis) Modification ML0713506512007-05-0808 May 2007 Slides, Category 2 Public Meeting with NEI on the Implications of the Wolf Creek Dissimilar Metal Weld Inspections. ML0713506532007-05-0808 May 2007 Slides, Recommendations for Critical Flaw Size Calculations (in Wolf Creek Advanced Fea Project). ML0713603732007-05-0101 May 2007 Enclosure-1 05/01/2007 Dominion Engineering Presentation Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential Pwscc. ML0713603962007-05-0101 May 2007 Enclosure 2-05/01/2007 Engineering Mechanics Corporation of Columbus Presentation, NRC Welding Residual Stress Solutions as Generated by Battelle and Emc2. ML0701601892006-12-19019 December 2006 12/19/2006, Viewgraphs from Meeting with Wolf Creek Generating Station to Discuss License Renewal Process and Environmental Scoping ML0635603462006-11-30030 November 2006 November 30, 2006 NRC Presentation Slides: Wolf Creek Flaw Evaluation ML0635603582006-11-30030 November 2006 Industry Presentation Slides: MRP-139 Analysis Basis ML0632100802006-11-16016 November 2006 Industry Presentations: Nov. 16, 2006 Public Meeting ML0632100772006-11-16016 November 2006 NRC Presentation: NRC Perspective on Wolf Creek Inspection Results ML0622804372006-08-16016 August 2006 Handouts from Wolf Creek Nuclear Operating Corporation and from Union Electric Company for Meeting with NRC on the Main Steam Isolation Valve (MSIV) Operability Determination ML0618000292006-06-28028 June 2006 Handout for June 28, 2006, Meeting with Representatives of Wolf Creek Nuclear Operating Corporation for Wolf Creek Generating Station on the Main Steam and Feedwater Isolation System (Msfis) Controls Replacement Project ML0615106012006-05-23023 May 2006 EPRI HRA Users Group Review of Draft NUREG-1842 ML0617902332006-03-0808 March 2006 RIC 2006 Presentation - W3D - Maurice E. Dingler - GSI 191 ML0516705662005-06-16016 June 2005 Summary of Annual Performance Assessment Meeting with NRC Re Oversight Process and Safety Performance at Wolf Creek Generating Station 2023-10-16
[Table view] Category:Meeting Briefing Package/Handouts
MONTHYEARML23275A1712023-10-16016 October 2023 October 16, 2023, Licensee Pre-submittal Meeting Slides - License Amendment Request to Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505-A, Revision 2 for Wolf Creek Generating Station ML21333A1512021-11-29029 November 2021 Pre-Submittal Meeting Slides License Amendment Request Addressing Portable Lighting for Operator Manual Actions ML21179A0892021-07-13013 July 2021 Background Information July 13, 2021, Pre-Submittal Public Teleconference ML20308A7332020-11-18018 November 2020 Slides for GSI-191 Pre-Submittal Public Teleconference November 18, 2020 ML18299A0492018-10-30030 October 2018 Accident Analyses Methodology Transition LAR Public Meeting IR 05000482/20174072017-12-22022 December 2017 Summary of Closed Regulatory Conference to Discuss Wolf Creek, Unit 1, Security Inspection Report 05000482/2017407 ML16271A4822016-09-27027 September 2016 Summary of Regulatory Conference to Discuss Safety Significance of Wolf Creek Generating Station Emergency Generator Excitation Diode Apparent Violation ML16236A0952016-08-25025 August 2016 and Wolf Creek Generating Station - August 25, 2016, Class 1E Electrical Equipment Air Conditioning System Pre-Application Meeting ML16095A0782016-04-12012 April 2016 Slides for the WCGS Meeting Discussion 4/12/16 ML16095A0852016-04-12012 April 2016 Large Scale Head Loss Test Specification 4/12/16 ML16095A0802016-04-12012 April 2016 Large Scale Penetration Test Specification 4/12/16 ML16095A0922016-03-31031 March 2016 Large Scale Penetration and Head Loss Test Plan 4/12/16 ML14188C4812014-07-0707 July 2014 Summary of Annual Assessment Meeting with Wolf Creek Generating Station ML14134A1842014-05-14014 May 2014 Summary of Public Meeting to Discuss an Apparent Violation Identified at Wolf Creek Generating Station ML14038A3852014-02-0606 February 2014 1/22/2014 - Summary of Public Meeting to Discuss Corrective Actions Implemented to Address the Chilling Effect Letter ML13218A0212013-08-0707 August 2013 Licensee Slides, 08/07/13 Preapplication Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Long-Term Corrective Actions for Water Hammer Events in the Essential Service Water System at Wolf Creek ML13205A2142013-07-30030 July 2013 Operating Corporation - Slides for Pre-application Meeting on 7/30/13 - Core Design and Safety Analysis Methodology Transition License Amendment Request for Wolf Creek Generating Station ML12263A3622012-09-20020 September 2012 Pre-application Meeting Slide Core Design and Safety Analysis Methodology Transition License Amendment Request TAC No. ME9495) ML12191A1742012-07-0909 July 2012 6/25/2012 Summary of Public Meeting with Wolf Creek Nuclear Operating Corporation ML1110104292011-04-11011 April 2011 Summary of Annual Performance Assessment Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Wolf Creek Generating Station Performance for the NRC Inspection Period Ending December 31, 2010 ML1016204902010-06-16016 June 2010 Notice of Construction Reactor Oversight Process Category 2 Public Meeting Handout: Wolf Creek 2, Notice of Violation and Proposed Imposition of Civil Penalty, Dated 11/21/1984 ML1016901162010-06-16016 June 2010 6/16/2010 Construction Reactor Oversight Process Category 2 Public Meeting Handout: Meeting Slides (Enforcement Cases) ML1013404892010-05-14014 May 2010 Summary of Annual Performance Assessment Meeting with Wolf Creek Nuclear Operating Corporation to Discuss the Wolf Creek Generating Station Performance for the NRC Inspection Period from January 1 Through December 31, 2009 ML0933702032009-12-0303 December 2009 Summary of Public Meeting for Wolf Creek Generating Station ML0933801482009-11-20020 November 2009 Licensee Handouts from November 20, 2009, Public Meeting with Union Electric Company and Wolf Creek Nuclear Operating Company to Discuss GL 2004-02 Response Rai'S ML0913300232009-05-13013 May 2009 Meeting Slides, Successful Licensing of the Als Fpga Based Safety Related I&C Platform ML0912600692009-05-0606 May 2009 Lessons Learned Using Digital I&C Interim Staff Guidance Workshop ML0902102422009-01-15015 January 2009 Licensee Slides, January 15, 2009, Category 1 Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Changes to Post Fire Shutdown Unresolved Items Analysis Methods Pre-Application for Wolf Creek Generating Station ML0834700222008-12-11011 December 2008 Slides, Category 2 Public Meeting Digital Instrumentation and Control Steering Committee M080717, M080717-Commission Briefing Slides/Exhibits Briefing on Fire Protection2008-07-17017 July 2008 M080717-Commission Briefing Slides/Exhibits Briefing on Fire Protection ML0817706682008-06-25025 June 2008 Summary of Meeting with Wolf Creek Nuclear Operating Corporation to Discuss Several Initiatives That Were Being Implemented to Improve Plant Performance at Wolf Creek ML0803706102008-01-11011 January 2008 Westinghouse Electric Company LLC, LTR-CDME-08-2, Rev. 1 NP-Attachment, Meeting Handouts from the December 13, 2007 Meeting with Wolf Creek and NRR on H*/B*. ML0729204682007-10-29029 October 2007 09/25/2007, Summary of Meeting Between the U.S. Nuclear Regulatory Commission Staff and Strategic Teaming and Resource Sharing Representatives to Discuss License Renewal Activities ML0721501532007-08-0202 August 2007 NRC Staff Handout for August 2, 2007, Meeting with Wolf Creek Nuclear Operating Corporation ML0721102382007-07-24024 July 2007 Performanc Contracting, Inc.'S Powerpoint Slide, Proposed Wolf Creek/Callaway Test Configuration ML0719303772007-07-11011 July 2007 Handouts (NRC and Licensee) for Meeting with Representatives of Wolf Creek Nuclear Operating Corporation ML0718606902007-06-19019 June 2007 06/19-20/2007 Slides,Fabrication Records Review, from Category 2 Public Meeting Between the NRC Staff and the Expert Panel for the Wolf Creek Advanced Finite Element Analyses (Fea) ML0716203882007-06-0101 June 2007 06/01/07 - Presentation Material, Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential Pwscc. ML0733409182007-05-31031 May 2007 05/31/2007 Presentation by J. Cudsworth NRC Treatment of Issues Other than Category 2 Issues for License Renewal ML0713703642007-05-17017 May 2007 Handout for Meeting with Wolf Creek Nuclear Operating Corporation on the Licensee'S Application for the Main Steam and Feedwater Isolation System (Msfis) Modification ML0713506532007-05-0808 May 2007 Slides, Recommendations for Critical Flaw Size Calculations (in Wolf Creek Advanced Fea Project). ML0713506462007-05-0808 May 2007 Industry Slides, Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential PWSCC, from Status Meeting on Implications of Wolf Creek Dissimilar Metal Weld Inspections ML0713506512007-05-0808 May 2007 Slides, Category 2 Public Meeting with NEI on the Implications of the Wolf Creek Dissimilar Metal Weld Inspections. ML0713603732007-05-0101 May 2007 Enclosure-1 05/01/2007 Dominion Engineering Presentation Advanced Fea Crack Growth Calculations for Evaluation of PWR Pressurizer Nozzle Dissimilar Metal Weld Circumferential Pwscc. ML0713603962007-05-0101 May 2007 Enclosure 2-05/01/2007 Engineering Mechanics Corporation of Columbus Presentation, NRC Welding Residual Stress Solutions as Generated by Battelle and Emc2. ML0701704732007-01-19019 January 2007 12/19/2006 Summary of Public Meetings Related to the Review of the Wolf Creek Generating Station License Renewal Application ML0635603632006-12-20020 December 2006 12/01/2006 Industry Meeting Handout: Draft Report (December 2006) - Implications of Wolf Creek Pressurizer Butt Weld Indications Relative to Safety Assessment and Inspection Requirements ML0701601892006-12-19019 December 2006 12/19/2006, Viewgraphs from Meeting with Wolf Creek Generating Station to Discuss License Renewal Process and Environmental Scoping ML0635603582006-11-30030 November 2006 Industry Presentation Slides: MRP-139 Analysis Basis ML0635603462006-11-30030 November 2006 November 30, 2006 NRC Presentation Slides: Wolf Creek Flaw Evaluation 2023-10-16
[Table view] |
Text
WCNOC-NRC WCNOC NRC Pre-submittal Pre submittal Meeting Core Design and Safety Analysis Methodology Transition License Amendment Request July 30 30, 2013 Wolf Creek Nuclear Operating p g Corporation p
1
Meeting Agenda
- Meeting M ti PPurpose / Objectives Obj ti
- License Amendment Request (LAR) Content
- Transition to Westinghouse Analysis Methodologies
- RTS/ESFAS/LOP DG Start Instrumentation Setpoint Uncertaintyy Analysis y
Introductions
- Teleconference Attendees 3
WCNOC Ownership of Safety Analysis
- WCNOC O Ownership hi off ththe S Safety f t AAnalysis l i iis d demonstrated t t d through the strength and application of the Quality Assurance (QA) Program
- WCNOC technical oversight is provided by the QA processes and participation in NUPIC
- WCNOC QA
- T Transmittal itt l off Design D i Information I f ti - referenced f d tto the th source calculation, design basis document, or identified as a direct input
- Configuration Management updates, regulatory reviews, etc.
- WCNOC/Westinghouse (WEC) Interface
- On-site WEC staff responsible for core design and safety analysis
- On-site On site WEC staff maintains qualifications in both the WCNOC and WEC QA programs. Proper independence is always maintained between the QA programs. 4
License Amendment Request Content
- License Amendment Request (LAR) to revise the WCGS Technical Specifications (TSs) based on:
- Transition to Westinghouse core design and safety analysis methodologies
- Transition to Westinghouse Setpoint Methodology
- Full Scope Implementation of Alternative Source Term (AST) 5
License Amendment Request Content
- LAR is consistent with NEI 06-02, License Amendment Request (LAR) Guidelines template Attachments:
I - Evaluation II - Proposed p TS Changes g ((Mark-up) p)
III - Revised TS pages IV - Proposed TS Bases Changes (info only)
V - Proposed COLR Changes (info only) 6
License Amendment Request Content
(cont )
Enclosures:
I - WCAP 17658 NP - Transition Licensing Report WCAP-17658-NP II - WCAP-17746-P - Setpoint Methodology for WCGS III - WCAP-17746-NP IV - WCAP-17602-P - Setpoint Calculations for WCGS V - WCAP-17602-NP VI - Full Scope Implementation of Alternative Source Term VII - CD-ROM containing Meteorological Data VIII - Proprietary Information Affidavit for WCAP-17746-P IX - Proprietary Information Affidavit for WCAP-17602-P 7
Transition to Westinghouse Analysis Methodologies
- Core C D Designi and dS Safety f t Analysis A l i Methodology M th d l
- The Non-LOCA safety analyses were analyzed with g
Westinghouse, , NRC approved pp methods
- All of the Westinghouse Non-LOCA methods are applicable to the WCGS
- The Containment response analyses of record are not impacted
- The SBLOCA and LBLOCA analyses of record are not impacted
- The Core Design and Fuel Rod Design will be y
evaluated for each reload cycle 8
Transition to Westinghouse Analysis Methodologies
- Overview O i off Analysis A l i M Methodology th d l S Scope
- Non-LOCA Safety Analyses
was used for majority of the analyses
- Other Codes that were used:
- TWINKLE (WCAP-7979-P-A)
- FACTRAN (WCAP-7908-A)
- Non-LOCA Thermal-Hydraulics (T&H) Safety Analyses
- RETRAN-02 (WCAP-14882-P-A) and WCAP-10698-P-A were used d ffor th the Steam St Generator G t Tube T b Rupture R t Margin M i to Overfill and Input to the Dose analyses 9
Transition to Westinghouse Analysis Methodologies
- Overview O i off Analysis A l i M Methodology th d l S Scope ((cont.)
t)
- DNB Correlations used in the VIPRE-01 DNBR calculations
- The WRB-2 DNB correlation will continue to be used as the primary DNB correlation for the T&H analyses of fuel regions above the first mixing vane grid
- The ABB-NV DNB correlation was used for the T&H analyses of fuel regions below the first mixing vane grid
- The WLOP DNB correlation was used for the T&H analyses that are outside the range of applicability of the WRB-2 and ABB-NV DNB correlations 10
Transition to Westinghouse Analysis Methodologies
- Implementation off the Westinghouse Non-LOCA OC Safety Analysis Methodology resulted in five changes to the current WCGS TSs
- SLs 2.1.1, Added the ABB-NV and WLOB DNB Correlations
- TS 3.3.1, RTS Function 10, Reactor Coolant Flow- Low
- TS 3.4.1, RCS Pressure, Temperature and Flow DNB Limits
- The Minimum Measured Flow (MMF) was relocated to the COLR and revised from 371,000 gpm to 376,000 gpm
- The Thermal Design Flow - 361,200 gpm, replaces the MMF 11
Transition to Westinghouse Analysis Methodologies
- IImplementation l t ti off theth Westinghouse W ti h Non-LOCA N LOCA Safety S f t Analysis Methodology results in five changes to the current WCGS TSs (cont.) ( )
- TS Table 3.7.1-1, OPERABLE MSSVs versus Maximum Allowable Power, the maximum allowable power for 4, 3, and 2 OPERABLE MSSVs was revised
- TS 5.6.5, COLR
- Added WCAP-9272-P-A, the Westinghouse Reload M th d l Methodology tto S Specification ifi ti 5 5.6.5 65b b.
- Deleted the WCNOC methodologies from Specification 5.6.5 b.
Transition to Westinghouse Analysis Methodologies
- Followup F ll ffrom th the 9/20/12 P Pre-Submittal S b itt l M Meeting ti
- Provide a roadmap of which code was used to analyze each postulated accident - Attachment I of the LAR includes a roadmap
- The Limitations, Restrictions and Conditions for the Westinghouse codes used in the Non-LOCA safety analyses are addressed in detail, including justifications in the LAR (Enclosure I, Appendix A) 13
Instrumentation Setpoint Uncertainty Analysis
- Uncertainty U t i t Analysis A l i
- Transition from the existing WCNOC Setpoint Methodology gy to the current Westinghouse g Setpoint p
Methodology as applied to WCGS for RTS, ESFAS and LOP DG Start Instrumentation (WCAP-17746-P,, Enclosure II of LAR))
- Technical Specification Changes
- TS 3.3.1,TS 3.3.2, and TS 3.3.5 Allowable Values were replaced with a Nominal Trip Setpoint
- TS Table 3.3.1-1, Overtemperature T, Note 1 and Overpower T, Note 2 14
Instrumentation Setpoint Uncertainty Analysis
- Uncertainty U t i t Analysis A l i ((cont.) t)
- Calculations were performed for the RTS, ESFAS, and LOP DG Start instrumentation Functions using the current Westinghouse setpoint methodology (WCAP-17602-P, Enclosure IV of LAR))
- Implementation of the Westinghouse Setpoint Methodology resulted in two changes to the existing WCGS Trip Setpoints
Instrumentation Setpoint Uncertainty Analysis
- TSTF TSTF-493-A, 493 A Revision R i i 4, 4 Option O ti A - Technical T h i l Specification Changes
- Variation from Option A - Nominal Trip Setpoint specified in the single column format based on the Westinghouse Setpoint Methodology
- TS changes h include i l d the th addition dditi off individual i di id l Surveillance Requirement footnotes to the applicable instrumentation Functions in accordance with Option p
A of TSTF-493, Revision 4 16
Instrumentation Setpoint Uncertainty Analysis
- Followup from the 9/20/12 Pre-Submittal Meeting
- The level of detail of the setpoint methodology and setpoint calculations for WCGS is consistent with that in the Diablo Canyon y Power Plant ((DCPP))
submittal of March 7, 2013 (DCL-13-016)
- WCNOC specific setpoint calculations are provided in WCAP WCAP-17602-P 17602 P (Enclosure IV of LAR) 17
Alternative Source Term
- Full F ll S Scope IImplementation l t ti off the th AST
- Radiological dose consequences analyses were performed for the accidents specified p p in Regulatory g y Guide (RG) 1.183 include:
- Main Steamline Break (USAR Section 15.1.5.3)
- Locked Rotor (USAR Section 15.3.3.3) 15 3 3 3)
- Rod Ejection (USAR Section 15.4.8.3)
- Loss of Coolant Accident (USAR Section 15.6.5.4)
- Fuel Handling Accident (USAR Section 15.7.4) 18
Alternative Source Term
- Full Scope Implementation of the AST (cont.)
- Radiological dose consequences analyses performed for additional accidents not specified in RG 1.183 include:
- Loss of Non-Emergency AC Power (USAR Section 15 2 6 3) 15.2.6.3)
- Letdown Line Break (USAR Section 15.6.2.1)
- Waste Gas Decay Tank Failure (USAR Section 15.7.1)
- Liquid Liq id Waste Tank Fail Failure re (USAR Section 15.7.2) 15 7 2)
- Dose consequences analyses were performed using version 3.03 of the RADTRAD computer code d
19
Alternative Source Term
- Full Scope Implementation of AST (cont.)
- No changes to the licensing basis EQ dose analyses - maintaining the TID TID-14844 14844 accident source term
- No changes to the licensing basis NUREG-0737 evaluations other than the Control Room Habitability Envelope (CHRE) doses (III.D.3.4) and Technical Support Center doses (III (III.A.1.2)
A 1 2) 20
Alternative Source Term
- Atmospheric Dispersion Factors (X/Q)
- New X/Q values were calculated
- Offsite (EAB and LPZ) X/Q values were calculated using the PAVAN code consistent with RG 1.145
- Control Room and TSC X/Q values were calculated using the ARCON96 code consistent with RG 1 1.194 194
- Meteorological Data
- Five years of WCGS site-specific meteorological data from 1/1/2006 through 12/31/2010 was collected
- Data recovery for the 5-year period met the 90% recovery criterion of RG 1.23 21
Alternative Source Term
- Current licensing basis changes
- Revises USAR Chapter 15 dose analysis for 10 accidents
((includes the 6 DBAs in RG 1.183))
- New Offsite, Control Room, and TSC atmospheric dispersion factors based on site-specific meteorological data from 2006 throughg 2010
- Revises the CRHE unfiltered inleakage from 20 scfm to 50 scfm
- Revises the Control Building unfiltered inleakage from 300 scfm to 400 scfm
- TS changes to address the update of the accident source term and associated DBAs
- TS changes to address the adoption of TSTF TSTF-51-A, 51 A Revision 2 22
Alternative Source Term
- Technical T h i lS Specification ifi ti Ch Changes
- Definition of DOSE EQUIVALENT I-131
- Revised to only allow the use of the dose conversion factors from EPA Federal Guidance Report No. 11
- Definition of DOSE EQUIVALENT XE-133
- Revised to only allow the use of the dose conversion factors EPA Federal Guidance Report No. 12
- Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring ProgramProgram
- Revises the quantity of radioactivity contained in each gas storage tank to be less than the amount that would result in a whole bodyy exposure p limit to 0.1 rem ((current limit is 0.5 rem) 23
Alternative Source Term
- Technical T h i lS Specification ifi ti Ch Changes ((cont.) t)
- Adoption of TSTF-51-A, Revision 2, Revise q
Containment Requirements during g Handling g Irradiated Fuel and Core Alterations
- Allows the elimination of the TS requirements for certain g
Engineered Safetyy Feature ((ESF)) systems y to be OPERABLE, after a sufficient radioactive decay has occurred
- Changes were not applied to the TS Section 3.8 Electrical TSs (conservative) 24
Alternative Source Term
- E l Enclosure VI off LAR - Sections S ti 1 - DESCRIPTION 2 - PROPOSED CHANGES 3 - BACKGROUND 4 - TECHNICAL ANALYSIS 5 - RG 1.183 CONFORMANCE TABLE 6 - RG 1.145 CONFORMANCE TABLE 7 - RG 1.194 CONFORMANCE TABLE 8 - RIS 2006-04 TABLE 9 - PROPOSED TS MARKUPS 10 - RETYPED TS PAGES 11 - PROPOSED BASES MARKUPS (information only) 12 - PROPOSED TRM and BASES MARKUP (information only) 25 13 - PROPOSED USAR CHANGES (information only)
Alternative Source Term
- Followup from the 9/20/12 Pre-Submittal Meeting
- Meteorological Data - One gap in the recorded data was due to the data logger failure (5/30/2007 through 6/7/2007)
- Provide a detailed plant drawing that shows the potential release paths - site plan provided consistent with the guidance in RIS 2006-04 (Enclosure VI of the LAR) 26
Alternative Source Term
- Followup from the 9/20/12 Pre-Submittal Meeting (cont) 27
Schedule
- Submit LAR to NRC August 13, 2013
- Requested Approval Date December 15, 2014
- Start of Refueling Outage 20 January 5, 2015
- Cycle 21 Startup February 9, 2015 28
Summary
- WCNOC iintends t d tto submit b it a LAR on 8/13/13 tto revise i
the WCGS TSs based on:
- Transition to the Westinghouse core design and safety analysis methodologies
- Transition to the Westinghouse Setpoint Methodology
- Full Scope Implementation of the Alternative Source Term (AST)
- Request NRC approval by 12/15/14 to support Cycle 21 operation (Feb. 2015)