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| {{#Wiki_filter:W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015Cynthia R. Hafenstine Manager Regulatory AffairsRA 15-0080U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555 | | {{#Wiki_filter:W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0080 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 |
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| ==Reference:== | | ==Reference:== |
| : 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOCto USNRC2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf CreekGenerating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR),"to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015 | | : 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOC to USNRC 2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf Creek Generating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015 |
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| ==Subject:== | | ==Subject:== |
| | | Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes Gentlemen: |
| Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency CoreCooling System (ECCS) Model ChangesGentlemen: | | In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM). |
| In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems forlight-water nuclear power reactors," | | The license amendment was implemented at WCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0 F in the PCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria set forth in 10 CFR 50.46. Additional reanalysis is not required.Attachment I provides an assessment of the specific changes to the Westinghouse ECCS evaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080 Page 2 of 2 Attachment II provides an update of the WCGS POT margin utilization for the large break LOCA evaluation model.This letter contains no commitments. |
| paragraph (a)(3)(ii), | | If you have any questions concerning this matter, please contact me at (620) 364-4204.Sincerely, Cynthia R. Hafenstine CRH/rlt Attachment I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Form cc: M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), wla Senior Resident Inspector (NRC), w/a Attachment I to RA 15-0080 Page 1 of 2 Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE |
| Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM). | |
| The license amendment was implemented atWCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0F in thePCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria setforth in 10 CFR 50.46. Additional reanalysis is not required. | |
| Attachment I provides an assessment of the specific changes to the Westinghouse ECCSevaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080Page 2 of 2Attachment II provides an update of the WCGS POT margin utilization for the large break LOCAevaluation model.This letter contains no commitments. | |
| If you have any questions concerning this matter, pleasecontact me at (620) 364-4204. | |
| Sincerely, Cynthia R. Hafenstine CRH/rltAttachment I Assessment of Changes to the Westinghouse Emergency Core CoolingSystem (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak CladdingTemperature (PCT) Margin Utilization Rack-up Formcc: M. L. Dapas (NRC), w/aC. F. Lyon (NRC), w/aN. H. Taylor (NRC), wlaSenior Resident Inspector (NRC), w/a Attachment I to RA 15-0080Page 1 of 2Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE | |
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| ===Background=== | | ===Background=== |
| Various changes have been made to enhance the usability of codes and to streamline futureanalyses. | | Various changes have been made to enhance the usability of codes and to streamline future analyses. |
| Examples of these changes include modifying input variable definitions, units anddefaults; | | Examples of these changes include modifying input variable definitions, units and defaults; ,improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451"Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0°0 F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORS Background Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. |
| ,improving the input diagnostic checks; enhancing the code output; optimizing activecoding; and eliminating inactive coding. These changes represent Discretionary Changes thatwill be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451 "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." | | The uncertainty factors for 2 3 9 Pu were applied to 2 3 8 U, and those for 2 3 8 U were applied to 2 3 9 pu. This error causes an over-prediction of the uncertainty in decay power from 2 3 9 pu and an under-prediction of the uncertainty in decay power from 2 3 8 U. Further, the decay group uncertainty factor for Decay Group 6 of 2 3 5 U was erroneously coded as 2.5% instead of 2.25%. Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. The resolution of these issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The issues described above were evaluated to account for the correction of these errors. The plant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf Creek Generating Station (WCGS). |
| Affected Evaluation Model2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe nature of these changes leads to an estimated peak cladding temperature (PCT) impact of0°0F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORSBackground Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. Thedecay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. | | Attachment I to PA 15-0080 Page 2 of 2 WCGS CONTAINMENT COOLING CAPACITY Background Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fan cooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis.This issue has been evaluated to estimate the impact on existing PCT results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The estimated effect was determined for the large break LOCA evaluation model based on the change in calculated containment, pressure resulting from the correct containment cooling capacity. |
| The uncertainty factors for 239Pu were applied to 238U, and those for 238U wereapplied to 239pu. This error causes an over-prediction of the uncertainty in decay power from239pu and an under-prediction of the uncertainty in decay power from 238U. Further, the decaygroup uncertainty factor for Decay Group 6 of 235U was erroneously coded as 2.5% instead of2.25%. Correction of these errors impacts the application of the sampled decay heatuncertainty, which may result in small changes to the decay heat power. These issues havebeen evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. | | The change in calculated containment pressure leads to an estimated effect of 0°F for the ASTRUM evaluation model analysis. |
| The resolution of theseissues represents a closely-related group of Non-Discretionary Changes in accordance withSection 4.1.2 of WCAP-1 3451.Affected Evaluation Model2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe issues described above were evaluated to account for the correction of these errors. Theplant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf CreekGenerating Station (WCGS). | | Attachment II to IRA 15-0080 Page 1 of 1 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION |
| Attachment I to PA 15-0080Page 2 of 2WCGS CONTAINMENT COOLING CAPACITYBackground Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fancooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis. | | ***Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting Break Size: LICENSING BASIS ASTRUM (2004)RFA-2 FQ=2.50, FdH=1.65 10%3565 MWth DEG Clad Temp (0 F)Ref. Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (APCT)1900°F 1 A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS |
| This issue has been evaluated to estimate the impact on existing PCT results. | | : 1. None B. PLANNED PLANT CHANGE EVALUATIONS |
| The resolution ofthis issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe estimated effect was determined for the large break LOCA evaluation model based on thechange in calculated containment, pressure resulting from the correct containment coolingcapacity. | | : 1. Containment Fan Cooler Capacity C. 2014 PERMANENT ECCS MODEL ASSESSMENTS |
| The change in calculated containment pressure leads to an estimated effect of 0°F forthe ASTRUM evaluation model analysis. | | : 1. Containment Fan Cooler Capacity 2. Decay Group Uncertainty Factors Errors D. OTHER 1. None 0 0 0-10 2 (a)2 3 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0 F |
| Attachment II to IRA 15-0080Page 1 of 1EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDINGTEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION | |
| ***Evaluation Model:Fuel:Peaking Factor:SG Tube Plugging: | |
| Power Level:Limiting Break Size:LICENSING BASISASTRUM (2004)RFA-2FQ=2.50, FdH=1.6510%3565 MWthDEGClad Temp (0F)Ref. NotesAnalysis of Record (AOR) PCTMARGIN ALLOCATIONS (APCT)1900°F 1A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS | |
| : 1. NoneB. PLANNED PLANT CHANGE EVALUATIONS | |
| : 1. Containment Fan Cooler CapacityC. 2014 PERMANENT ECCS MODEL ASSESSMENTS | |
| : 1. Containment Fan Cooler Capacity2. Decay Group Uncertainty Factors ErrorsD. OTHER1. None000-102 (a)230LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0F | |
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| ==References:== | | ==References:== |
| : 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," | | : 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.Notes: (a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers. |
| January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," | | W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0080 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 |
| August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the DecayGroup Uncertainty Factors Errors," | |
| November 2014.Notes:(a) This effect was estimated based on a cooling capacity intended to bound futureimplementation of replacement tube bundles in the containment fan coolers. | |
| W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015Cynthia R. Hafenstine Manager Regulatory AffairsRA 15-0080U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555 | |
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| ==Reference:== | | ==Reference:== |
| : 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOCto USNRC2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf CreekGenerating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR),"to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015 | | : 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOC to USNRC 2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf Creek Generating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015 |
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| ==Subject:== | | ==Subject:== |
| | | Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes Gentlemen: |
| Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency CoreCooling System (ECCS) Model ChangesGentlemen: | | In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM). |
| In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems forlight-water nuclear power reactors," | | The license amendment was implemented at WCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0 F in the PCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria set forth in 10 CFR 50.46. Additional reanalysis is not required.Attachment I provides an assessment of the specific changes to the Westinghouse ECCS evaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080 Page 2 of 2 Attachment II provides an update of the WCGS POT margin utilization for the large break LOCA evaluation model.This letter contains no commitments. |
| paragraph (a)(3)(ii), | | If you have any questions concerning this matter, please contact me at (620) 364-4204.Sincerely, Cynthia R. Hafenstine CRH/rlt Attachment I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Form cc: M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), wla Senior Resident Inspector (NRC), w/a Attachment I to RA 15-0080 Page 1 of 2 Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE |
| Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM). | |
| The license amendment was implemented atWCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0F in thePCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria setforth in 10 CFR 50.46. Additional reanalysis is not required. | |
| Attachment I provides an assessment of the specific changes to the Westinghouse ECCSevaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080Page 2 of 2Attachment II provides an update of the WCGS POT margin utilization for the large break LOCAevaluation model.This letter contains no commitments. | |
| If you have any questions concerning this matter, pleasecontact me at (620) 364-4204. | |
| Sincerely, Cynthia R. Hafenstine CRH/rltAttachment I Assessment of Changes to the Westinghouse Emergency Core CoolingSystem (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak CladdingTemperature (PCT) Margin Utilization Rack-up Formcc: M. L. Dapas (NRC), w/aC. F. Lyon (NRC), w/aN. H. Taylor (NRC), wlaSenior Resident Inspector (NRC), w/a Attachment I to RA 15-0080Page 1 of 2Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE | |
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| ===Background=== | | ===Background=== |
| Various changes have been made to enhance the usability of codes and to streamline futureanalyses. | | Various changes have been made to enhance the usability of codes and to streamline future analyses. |
| Examples of these changes include modifying input variable definitions, units anddefaults; | | Examples of these changes include modifying input variable definitions, units and defaults; ,improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451"Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0°0 F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORS Background Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. |
| ,improving the input diagnostic checks; enhancing the code output; optimizing activecoding; and eliminating inactive coding. These changes represent Discretionary Changes thatwill be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451 "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." | | The uncertainty factors for 2 3 9 Pu were applied to 2 3 8 U, and those for 2 3 8 U were applied to 2 3 9 pu. This error causes an over-prediction of the uncertainty in decay power from 2 3 9 pu and an under-prediction of the uncertainty in decay power from 2 3 8 U. Further, the decay group uncertainty factor for Decay Group 6 of 2 3 5 U was erroneously coded as 2.5% instead of 2.25%. Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. The resolution of these issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The issues described above were evaluated to account for the correction of these errors. The plant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf Creek Generating Station (WCGS). |
| Affected Evaluation Model2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe nature of these changes leads to an estimated peak cladding temperature (PCT) impact of0°0F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORSBackground Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. Thedecay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. | | Attachment I to PA 15-0080 Page 2 of 2 WCGS CONTAINMENT COOLING CAPACITY Background Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fan cooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis.This issue has been evaluated to estimate the impact on existing PCT results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The estimated effect was determined for the large break LOCA evaluation model based on the change in calculated containment, pressure resulting from the correct containment cooling capacity. |
| The uncertainty factors for 239Pu were applied to 238U, and those for 238U wereapplied to 239pu. This error causes an over-prediction of the uncertainty in decay power from239pu and an under-prediction of the uncertainty in decay power from 238U. Further, the decaygroup uncertainty factor for Decay Group 6 of 235U was erroneously coded as 2.5% instead of2.25%. Correction of these errors impacts the application of the sampled decay heatuncertainty, which may result in small changes to the decay heat power. These issues havebeen evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. | | The change in calculated containment pressure leads to an estimated effect of 0°F for the ASTRUM evaluation model analysis. |
| The resolution of theseissues represents a closely-related group of Non-Discretionary Changes in accordance withSection 4.1.2 of WCAP-1 3451.Affected Evaluation Model2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe issues described above were evaluated to account for the correction of these errors. Theplant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf CreekGenerating Station (WCGS). | | Attachment II to IRA 15-0080 Page 1 of 1 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION |
| Attachment I to PA 15-0080Page 2 of 2WCGS CONTAINMENT COOLING CAPACITYBackground Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fancooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis. | | ***Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting Break Size: LICENSING BASIS ASTRUM (2004)RFA-2 FQ=2.50, FdH=1.65 10%3565 MWth DEG Clad Temp (0 F)Ref. Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (APCT)1900°F 1 A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS |
| This issue has been evaluated to estimate the impact on existing PCT results. | | : 1. None B. PLANNED PLANT CHANGE EVALUATIONS |
| The resolution ofthis issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUMEstimated EffectThe estimated effect was determined for the large break LOCA evaluation model based on thechange in calculated containment, pressure resulting from the correct containment coolingcapacity. | | : 1. Containment Fan Cooler Capacity C. 2014 PERMANENT ECCS MODEL ASSESSMENTS |
| The change in calculated containment pressure leads to an estimated effect of 0°F forthe ASTRUM evaluation model analysis. | | : 1. Containment Fan Cooler Capacity 2. Decay Group Uncertainty Factors Errors D. OTHER 1. None 0 0 0-10 2 (a)2 3 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0 F |
| Attachment II to IRA 15-0080Page 1 of 1EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDINGTEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION | |
| ***Evaluation Model:Fuel:Peaking Factor:SG Tube Plugging: | |
| Power Level:Limiting Break Size:LICENSING BASISASTRUM (2004)RFA-2FQ=2.50, FdH=1.6510%3565 MWthDEGClad Temp (0F)Ref. NotesAnalysis of Record (AOR) PCTMARGIN ALLOCATIONS (APCT)1900°F 1A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS | |
| : 1. NoneB. PLANNED PLANT CHANGE EVALUATIONS | |
| : 1. Containment Fan Cooler CapacityC. 2014 PERMANENT ECCS MODEL ASSESSMENTS | |
| : 1. Containment Fan Cooler Capacity2. Decay Group Uncertainty Factors ErrorsD. OTHER1. None000-102 (a)230LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0F | |
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| ==References:== | | ==References:== |
| : 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," | | : 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.Notes: (a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.}} |
| January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," | |
| August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the DecayGroup Uncertainty Factors Errors," | |
| November 2014.Notes:(a) This effect was estimated based on a cooling capacity intended to bound futureimplementation of replacement tube bundles in the containment fan coolers.}} | |
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Category:Letter
MONTHYEARIR 05000482/20244202024-10-31031 October 2024
Security Baseline Inspection Report 05000482/2024420
ML24296B1902024-10-22022 October 2024
10 CFR 50.55a Requests for the Fifth Ten-Year Interval Inservice Testing Program
05000482/LER-2024-001-01, Mode 3 Entry with One Auxiliary Feedwater Pump Train Inoperable Due to Missed Post-Maintenance Testing2024-10-22022 October 2024
Mode 3 Entry with One Auxiliary Feedwater Pump Train Inoperable Due to Missed Post-Maintenance Testing
05000482/LER-2024-002, Technical Specification Required Shutdown Due to Inoperable Auxiliary Feedwater Discharge Valve2024-10-21021 October 2024
Technical Specification Required Shutdown Due to Inoperable Auxiliary Feedwater Discharge Valve
ML24284A2842024-10-10010 October 2024
(Wcgs), Revision of One Form That Implements the Radiological Emergency Response Plan (RERP)
ML24283A0752024-10-0909 October 2024
Notification of Commercial Grade Dedication Inspection (05000482/2025012) and Request for Information
ML24199A1712024-09-17017 September 2024
Issuance of Amendment No. 241 Revise Ventilation Filter Testing Program Criteria and Administrative Correction of Absorber in Technical Specification 5.5.11
ML24260A0712024-09-12012 September 2024
License Amendment Request for a Risk-Informed Resolution to GSI-191
IR 05000482/20240102024-09-10010 September 2024
Biennial Problem Identification and Resolution Inspection Report 05000482/2024010 (Public)
ML24255A8642024-09-0606 September 2024
Rscc Wire & Cable LLC Dba Marmon Industrial Energy & Infrastructure - Part 21 Retraction of Final Notification
ML24248A0762024-09-0404 September 2024
Containment Inservice Inspection Program Third Interval, Second Period, Refueling Outage 26 Owner’S Activity Report
ML24248A2492024-09-0404 September 2024
Inservice Inspection Program Fourth Interval, Third Period, Refueling Outage 26 Owner’S Activity Report
ML24241A2212024-08-29029 August 2024
Notice of Enforcement Discretion for Wolf Creek Generating Station
ML24240A2642024-08-27027 August 2024
Corporation - Request for Notice of Enforcement Discretion Regarding Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System
ML24239A3972024-08-23023 August 2024
Rssc Wire & Cable LLC Dba Marmon - Part 21 Final Notification - 57243-EN 57243
IR 05000482/20240052024-08-14014 August 2024
Updated Inspection Plan for Wolf Creek Generating Station (Report 05000482/2024005)
ML24227A5562024-08-14014 August 2024
Application to Revise Technical Specifications to Adopt TSTF-569-A, Revision 2, Revision of Response Time Testing Definitions
ML24213A3352024-07-31031 July 2024
License Amendment Request to Revise Technical Specification 3.2.1, Heat Flux Hot Channel Factor (Fq(Z)) (Fq Methodology), to Implement the Methodology from WCAP-17661-P-A, Revision 1.
ML24206A1252024-07-24024 July 2024
Revision of Three Procedures and Two Forms That Implement the Radiological Emergency Response Plan (RERP)
IR 05000482/20240022024-07-18018 July 2024
Integrated Inspection Report 05000482/2024002
05000482/LER-2024-001, Mode 3 Entry with One Auxiliary Feedwater Pump Train Inoperable Due to Missed Post-Maintenance Testing2024-07-0202 July 2024
Mode 3 Entry with One Auxiliary Feedwater Pump Train Inoperable Due to Missed Post-Maintenance Testing
IR 05000482/20244012024-07-0202 July 2024
Security Baseline Inspection Report 05000482/2024401
ML24178A3672024-06-26026 June 2024
Correction to 2023 Annual Radioactive Effluent Release Report – Report 47
ML24178A4142024-06-26026 June 2024
Revision of One Procedure and One Form That Implement the Radiological Emergency Response Plan (RERP)
ML24162A1632024-06-11011 June 2024
Operating Corporation – Notification of Biennial Problem Identification and Resolution Inspection and Request for Information (05000482/2024010)
ML24150A0562024-05-29029 May 2024
Foreign Ownership, Control or Influence (FOCI) Information – Change to Lists of Owners, Officers, Directors and Executive Personnel - Form 405F Amendment
ML23345A1602024-05-0909 May 2024
Revision of Safety Evaluation for Amendment No. 237 Request for Deviation from Fire Protection Requirements
ML24089A2622024-04-29029 April 2024
Financial Protection Levels
ML24118A0022024-04-27027 April 2024
Wolf Generating Nuclear Station - 2023 Annual Radiological Environmental Operating Report
ML24118A0032024-04-27027 April 2024
2023 Annual Radioactive Effluent Release Report - Report 47
ML24113A1882024-04-19019 April 2024
Foreign Ownership, Control or Influence Information - Change to Lists of Owners, Officers, Directors and Executive Personnel - Form 405F Amendment
ML24109A1842024-04-18018 April 2024
Cycle 27 Core Operating Limits Report
ML24109A1212024-04-18018 April 2024
(WCGS) Form 5 Exposure Report for Calendar Year 2023
IR 05000482/20240012024-04-17017 April 2024
Integrated Inspection Report 05000482/2024001
ML24114A1442024-04-15015 April 2024
Redacted Updated Safety Analysis Report (WCGS Usar), Revision 37
ML24106A1482024-04-15015 April 2024
Notification of Inspection (NRC Inspection Report 05000482/2024003) and Request for Information
ML24098A0052024-04-0707 April 2024
2023 Annual Environmental Operating Report
ML24089A0972024-03-29029 March 2024
Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations
ML24089A1352024-03-29029 March 2024
10 CFR 50.46 Annual Report of Emergency Core Cooling System (ECCS) Evaluation Model Changes
ML24074A3312024-03-14014 March 2024
Missed Quarterly Inspection Per 40 CFR 266 Subpart N
IR 05000482/20240122024-03-11011 March 2024
Fire Protection Team Inspection Report 05000482/2024012
ML24080A3452024-03-11011 March 2024
7 of the Wolf Creek Generating Station Updated Safety Analysis Report
ML24016A0702024-03-0808 March 2024
Issuance of Amendment No. 240 Removal of the Power Range Neutron Flux Rate - High Negative Rate Trip Function from Technical Specifications
ML24068A1992024-03-0707 March 2024
Changes to Technical Specification Bases - Revisions 93 and 94
ML24066A0672024-03-0505 March 2024
4-2022-024 Letter - OI Closure to Licensee
ML24061A2642024-03-0101 March 2024
Revision of Two Procedures That Implement the Radiological Emergency Response Plan (RERP) for Wolf Creek Generating Station (WCGS) Commissioners
IR 05000482/20230062024-02-28028 February 2024
Annual Assessment Letter for Wolf Creek Generating Station Report 05000482/2023006
ML24059A1702024-02-28028 February 2024
Annual Fitness for Duty Program Performance Report, and Annual Fatigue Report for 2023
ML24026A0212024-02-27027 February 2024
Issuance of Amendment No. 239 Modified Implementation Date of License Amendment No. 238
ML24050A0012024-02-19019 February 2024
(Wcgs), Revision of One Form That Implements the Radiological Emergency Response Plan (RERP)
2024-09-06
[Table view]
Category:Report
MONTHYEARML24248A0762024-09-0404 September 2024
Containment Inservice Inspection Program Third Interval, Second Period, Refueling Outage 26 Owner’S Activity Report
ML23334A2502023-11-30030 November 2023
Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation
ML23292A3572023-10-19019 October 2023
Operating Corporation, Request for Correction in Response to Issuance of Amendment No. 237 and Corresponding Safety Evaluation
ML23244A2142023-09-0101 September 2023
IAEA Osart Report
ML23068A1962023-03-0909 March 2023
Biennial 50.59 Evaluation Report
ML23135A2012022-12-0606 December 2022
EQSD-I, Rev 18, Equipment Qualification Design Basis Document
ML23135A1932022-09-0101 September 2022
EQSD-II, Rev 34, Equipment Qualification Summary Document
WO 22-0021, Technical Specification 5.6.8 - Post Accident Monitoring (PAM) Report2022-08-29029 August 2022
Technical Specification 5.6.8 - Post Accident Monitoring (PAM) Report
ML22152A2042022-05-18018 May 2022
EQSD-I, Revision 16, Equipment Qualification Design Basis Document
ML22152A1992022-05-18018 May 2022
EQSD-II, Revision 33, EQ Program Master List (Table 1) Plus Mild Environment SR Electrical Equipment List (Table 2)
ML22152A1942022-05-18018 May 2022
EQSD-I, Revision 15, Equipment Qualification Design Basis Document
ML23135A2072022-04-0606 April 2022
EQSD-I, Rev 17, Equipment Qualification Design Basis Document
ML20290A6192020-07-23023 July 2020
Corrected Redacted Version of Rev. 33 to Updated Safety Analysis Report, EQSD-I, Rev. 14, Equipment Qualification Design Basis Document
ML20182A4292020-06-30030 June 2020
Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating
WO 20-0048, Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating2020-06-30030 June 2020
Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating
WO 20-0041, Rev. 32 to Equipment Qualification Summary Document-II (EQSD-II)2020-01-15015 January 2020
Rev. 32 to Equipment Qualification Summary Document-II (EQSD-II)
ML19330F8632019-12-0606 December 2019
Regulatory Audit Summary Regarding License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation,
ML19178A3392019-07-0909 July 2019
Staff Assessment of Aging Management Program and Inspection Plan of Reactor Vessel Internals
ML19092A1562018-09-18018 September 2018
EQSD-I, Rev 14
ML18093B0112018-03-0808 March 2018
Revision 31 to Updated Safety Analysis Report, Equipment Classification Design Basis Document, EQSD-I Rev 13
ML18093B0132018-03-0808 March 2018
Revision 31 to Updated Safety Analysis Report, Equipment Qualification Summary Document, EQSD-II, Rev 30, Redacted
WO 19-0010, EQSD-II, Rev 312018-02-19019 February 2018
EQSD-II, Rev 31
ML18023A1572018-01-18018 January 2018
Submittal of Blind Sample Error Investigation Report
ML17241A2512017-11-0808 November 2017
Staff Assessment of Flooding Focused Evaluation
ML17234A4002017-08-17017 August 2017
Summary of Changes to Three Forms That Implement Radiological Emergency Response Plan (RERP)
ML17093A6652017-03-27027 March 2017
Report on Financial Protection Levels
ML17151B0452017-03-0909 March 2017
Equipment Qualification Summary Document, Master List Section II, EQSD-II, Rev 28, Part 2 - Redacted
ML17151B0442017-03-0202 March 2017
Equipment Qualification Summary Document, Master List Section II, EQSD-II, Rev 28, Part 1
ML17072A0512017-03-0202 March 2017
Biennial 50.59 Evaluation Report for 01/01/2015 - 12/31/2016
ML17151B0432017-01-26026 January 2017
Equipment Qualification Design Basis Document, EQSD-I, Rev 11
ML17054C2302017-01-17017 January 2017
License Amendment Request for the Transition to Westinghouse Core Design and Safety Analyses - A-3788, Technical Report, Appendix a, Fission Product Removal Effectiveness of Chemical Additives in PWR Containment Sprays
ET 17-0005, 16C4405-RPT-002, Revision 0, Seismic High Frequency Confirmation for WCGS2016-12-15015 December 2016
16C4405-RPT-002, Revision 0, Seismic High Frequency Confirmation for WCGS
WM 16-0008, Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, EQSD-I, Revision 9, Equipment Qualification Design Basis Document.2016-05-31031 May 2016
Wolf Creek Generating Station, Redacted Version of Revision 29 to Updated Safety Analysis Report, EQSD-I, Revision 9, Equipment Qualification Design Basis Document.
ML16204A1802016-05-31031 May 2016
Redacted Version of Revision 29 to Updated Safety Analysis Report, EQSD-I, Revision 9, Equipment Qualification Design Basis Document.
WM 16-0008, Redacted Version of Revision 29 to Updated Safety Analysis Report, EQSD-II, Revision 27 Redacted2016-05-31031 May 2016
Redacted Version of Revision 29 to Updated Safety Analysis Report, EQSD-II, Revision 27 Redacted
ML15314A6572015-11-0404 November 2015
Submittal of 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes
ML17151B0392015-10-26026 October 2015
E-1F9915, Rev 8, Design Basis Document for Ofn RP-017, Control Room Evaluation
ET 16-0003, Flood Hazard Reevaluation Report, Revision 12015-10-21021 October 2015
Flood Hazard Reevaluation Report, Revision 1
ML15216A3202015-08-12012 August 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights
ML15075A0282015-03-0404 March 2015
Biennial 50.59 Evaluation Report
ML15062A2752015-02-24024 February 2015
Pressure and Temperature Limits Report, Revision 2
ET 15-0002, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.6.5, for Large Break Loss-of-Coolant (LOCA) Analysis Methodology2015-01-21021 January 2015
Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.6.5, for Large Break Loss-of-Coolant (LOCA) Analysis Methodology
WO 14-0095, Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2014-12-23023 December 2014
Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
ML14189A5372014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5
ML14189A5432014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
ML14189A5332014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5
ML14189A5312014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5
ML14189A5002014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 5
ML14189A4982014-06-19019 June 2014
Stars Response to Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program - Vendor Information Request
2024-09-04
[Table view]
Category:Miscellaneous
MONTHYEARML24248A0762024-09-0404 September 2024
Containment Inservice Inspection Program Third Interval, Second Period, Refueling Outage 26 Owner’S Activity Report
ML23334A2502023-11-30030 November 2023
Request for Exemption from Enhanced Weapons, Firearms Background Checks, and Security Event Notifications Implementation
ML23292A3572023-10-19019 October 2023
Operating Corporation, Request for Correction in Response to Issuance of Amendment No. 237 and Corresponding Safety Evaluation
ML23068A1962023-03-0909 March 2023
Biennial 50.59 Evaluation Report
WO 20-0048, Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating2020-06-30030 June 2020
Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating
ML20182A4292020-06-30030 June 2020
Request for Notice of Enforcement Discretion for Technical Specifications 3.8.1, AC Sources - Operating
WO 20-0041, Rev. 32 to Equipment Qualification Summary Document-II (EQSD-II)2020-01-15015 January 2020
Rev. 32 to Equipment Qualification Summary Document-II (EQSD-II)
ML19330F8632019-12-0606 December 2019
Regulatory Audit Summary Regarding License Amendment Request to Revise Technical Specification 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation,
ML18023A1572018-01-18018 January 2018
Submittal of Blind Sample Error Investigation Report
ML17234A4002017-08-17017 August 2017
Summary of Changes to Three Forms That Implement Radiological Emergency Response Plan (RERP)
ML17093A6652017-03-27027 March 2017
Report on Financial Protection Levels
ET 17-0005, 16C4405-RPT-002, Revision 0, Seismic High Frequency Confirmation for WCGS2016-12-15015 December 2016
16C4405-RPT-002, Revision 0, Seismic High Frequency Confirmation for WCGS
ML15314A6572015-11-0404 November 2015
Submittal of 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes
ET 16-0003, Flood Hazard Reevaluation Report, Revision 12015-10-21021 October 2015
Flood Hazard Reevaluation Report, Revision 1
ML15216A3202015-08-12012 August 2015
Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights
ML15075A0282015-03-0404 March 2015
Biennial 50.59 Evaluation Report
ET 15-0002, Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.6.5, for Large Break Loss-of-Coolant (LOCA) Analysis Methodology2015-01-21021 January 2015
Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 5.6.5, for Large Break Loss-of-Coolant (LOCA) Analysis Methodology
WO 14-0095, Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2014-12-23023 December 2014
Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
ML14189A5432014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
ML14189A5312014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5
ML14189A5332014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5
ML14189A5372014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5
ML14189A5002014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 1 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 2 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 3 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
RIS 2014-07, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 52014-06-19019 June 2014
Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 4 of 5, Regulatory Issue Summary (RIS) 2014-07, Enhancements to the Vendor Inspection Program- Vendor Information Request. Part 5 of 5
ML14136A1712014-05-30030 May 2014
Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima Dai-Ichi Nuclear Power Plant Accident
ML14077A2812014-02-10010 February 2014
Operating Corporation Flood Hazard Reevaluation Report,Revision 0
ML1307206802013-03-0404 March 2013
Cycle 20 Core Operating Limits Report
ET 12-0032, WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-1 Through Page B-502012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-1 Through Page B-50
ML12342A2502012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-51 Through Page B-100
ML12342A2512012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-101 Through Page B-150
ML12342A2532012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Cover Through Appendix a, Seismic Walkdown Checklists (Swcs) Page A-112
ML12342A2542012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-151 Through Page B-184
ML12342A2552012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix B, Area Walk-by Checklists (Awcs), Page B-185 Through Page B-228
ML12342A2572012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix C, Licensing Basis Evaluations Summary, Page C-1 Through End
ML12342A2592012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix a, Seismic Walkdown Checklists (Swcs) Page A-113 Through Page A-222
ML12342A2602012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix a, Seismic Walkdown Checklists (Swcs) Page A-223 Through Page A-320
ML12342A2612012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix a, Seismic Walkdown Checklists (Swcs) Page A-321 Through Page A-350
ML12342A2622012-11-30030 November 2012
WCAP-17678-NP, Rev. 0, Wolf Creek Generating Station Post-Fukushima NTTF 2.3 Seismic Walkdown Submittal Report. Appendix a, Seismic Walkdown Checklists (Swcs) Page A-348 Through Page A-440
ML12100A0762012-03-26026 March 2012
Foreign Ownership, Control or Influence Information - Updated Lists of Owners, Officers, Directors and Executive Personnel
ML11133A3482011-05-0808 May 2011
Great Plains Energy, Inc. and Kansas Electric Power Cooperative, Inc. - Item 8. Financial Statemnts and Supplemental Data
ML11279A0052011-03-31031 March 2011
John Redmond Reservoir Diversion and Impacts to Downstream Conditions: System Simulation Modeling
ML1107500302011-03-10010 March 2011
Biennial 50.59 Evaluation Report
ML1023501702010-09-15015 September 2010
Review of the Results of 16th Steam Generator Tube Inservice Inspections Performed During Refueling Outage 17
ML0917607642009-06-24024 June 2009
Phase 3 Potential Loss of RHR Due to Inadequate Cooling of Suction Line During Mode 4 to Mode 3 Transition at Wolf Creek
ET 09-0010, Pressure and Temperature Limits Report, Revision 12009-03-0606 March 2009
Pressure and Temperature Limits Report, Revision 1
ET 09-0005, Submittal of Requested Information Regarding Main Steam and Feedwater Isolation System (Msfis) Controls Modification2009-01-29029 January 2009
Submittal of Requested Information Regarding Main Steam and Feedwater Isolation System (Msfis) Controls Modification
2024-09-04
[Table view] |
Text
W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0080 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Reference:
- 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOC to USNRC 2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf Creek Generating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015
Subject:
Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes Gentlemen:
In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM).
The license amendment was implemented at WCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0 F in the PCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria set forth in 10 CFR 50.46. Additional reanalysis is not required.Attachment I provides an assessment of the specific changes to the Westinghouse ECCS evaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080 Page 2 of 2 Attachment II provides an update of the WCGS POT margin utilization for the large break LOCA evaluation model.This letter contains no commitments.
If you have any questions concerning this matter, please contact me at (620) 364-4204.Sincerely, Cynthia R. Hafenstine CRH/rlt Attachment I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Form cc: M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), wla Senior Resident Inspector (NRC), w/a Attachment I to RA 15-0080 Page 1 of 2 Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses.
Examples of these changes include modifying input variable definitions, units and defaults; ,improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451"Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0°0 F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORS Background Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A.
The uncertainty factors for 2 3 9 Pu were applied to 2 3 8 U, and those for 2 3 8 U were applied to 2 3 9 pu. This error causes an over-prediction of the uncertainty in decay power from 2 3 9 pu and an under-prediction of the uncertainty in decay power from 2 3 8 U. Further, the decay group uncertainty factor for Decay Group 6 of 2 3 5 U was erroneously coded as 2.5% instead of 2.25%. Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. The resolution of these issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The issues described above were evaluated to account for the correction of these errors. The plant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf Creek Generating Station (WCGS).
Attachment I to PA 15-0080 Page 2 of 2 WCGS CONTAINMENT COOLING CAPACITY Background Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fan cooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis.This issue has been evaluated to estimate the impact on existing PCT results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The estimated effect was determined for the large break LOCA evaluation model based on the change in calculated containment, pressure resulting from the correct containment cooling capacity.
The change in calculated containment pressure leads to an estimated effect of 0°F for the ASTRUM evaluation model analysis.
Attachment II to IRA 15-0080 Page 1 of 1 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION
- Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting Break Size: LICENSING BASIS ASTRUM (2004)RFA-2 FQ=2.50, FdH=1.65 10%3565 MWth DEG Clad Temp (0 F)Ref. Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (APCT)1900°F 1 A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. None B. PLANNED PLANT CHANGE EVALUATIONS
- 1. Containment Fan Cooler Capacity C. 2014 PERMANENT ECCS MODEL ASSESSMENTS
- 1. Containment Fan Cooler Capacity 2. Decay Group Uncertainty Factors Errors D. OTHER 1. None 0 0 0-10 2 (a)2 3 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0 F
References:
- 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.Notes: (a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.
W~FCREEK'NUCLEAR OPERATING CORPORATION November 4, 2015 Cynthia R. Hafenstine Manager Regulatory Affairs RA 15-0080 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Reference:
- 1) Letter RA 15-0025, dated March 20, 2015, from S.R. Koenig, WCNOC to USNRC 2) Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf Creek Generating Station -Issuance of Amendment re: Revise Technical Specification 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)," to Add ASTRUM to the List of Analytical Methods (TAC NO.MF351 8)," dated August 28, 2015
Subject:
Docket No. 50-482: 10 CFR 50.46 Thirty Day Report of Emergency Core Cooling System (ECCS) Model Changes Gentlemen:
In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), Wolf Creek Nuclear Operating Corporation (WCNOC) is submitting the attached information to fulfill the 30-day reporting requirement for the Wolf Creek Generating Station (WCGS).In Reference 1, WCNOC reported the WCGS peak cladding temperature (PCT), calculated based on an-acceptable evaluation model. A new best-estimate large break loss-of-coolant (LOCA) methodology was approved for WCGS in Reference 2, known as Automated Statistical Treatment of Uncertainty Method (ASTRUM).
The license amendment was implemented at WCGS on October 26, 2015. The new analysis resulted in changes of greater than 50 0 F in the PCT from those previously reported to the NRC in the last 10 CFR 50.46 report (Reference 1).The calculated PCT for the WCGS large break LOCA remains within the acceptance criteria set forth in 10 CFR 50.46. Additional reanalysis is not required.Attachment I provides an assessment of the specific changes to the Westinghouse ECCS evaluation model for large break LOCAs.P.O. Box 41 IBurlington, KS 66839 I Phone: (620) 364-8831 P '-An Equal Opportunity Employer M/F/HCNVET RA 15-0080 Page 2 of 2 Attachment II provides an update of the WCGS POT margin utilization for the large break LOCA evaluation model.This letter contains no commitments.
If you have any questions concerning this matter, please contact me at (620) 364-4204.Sincerely, Cynthia R. Hafenstine CRH/rlt Attachment I Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS) Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)II Emergency Core Cooling System (ECCS) Evaluation Model Peak Cladding Temperature (PCT) Margin Utilization Rack-up Form cc: M. L. Dapas (NRC), w/a C. F. Lyon (NRC), w/a N. H. Taylor (NRC), wla Senior Resident Inspector (NRC), w/a Attachment I to RA 15-0080 Page 1 of 2 Assessment of Changes to the Westinghouse Emergency Core Cooling System (ECCS)Evaluation Model for Large Break Loss-of-Coolant Accident (LOCA)GENERAL CODE MAINTENANCE
Background
Various changes have been made to enhance the usability of codes and to streamline future analyses.
Examples of these changes include modifying input variable definitions, units and defaults; ,improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 of WCAP-13451"Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting." Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated peak cladding temperature (PCT) impact of 0°0 F.ERRORS IN DECAY GROUP UNCERTAINTY FACTORS Background Errors in the calculation of decay heat were discovered in the WCOBRA/TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A.
The uncertainty factors for 2 3 9 Pu were applied to 2 3 8 U, and those for 2 3 8 U were applied to 2 3 9 pu. This error causes an over-prediction of the uncertainty in decay power from 2 3 9 pu and an under-prediction of the uncertainty in decay power from 2 3 8 U. Further, the decay group uncertainty factor for Decay Group 6 of 2 3 5 U was erroneously coded as 2.5% instead of 2.25%. Correction of these errors impacts the application of the sampled decay heat uncertainty, which may result in small changes to the decay heat power. These issues have been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) best-estimate large break LOCA analysis results. The resolution of these issues represents a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The issues described above were evaluated to account for the correction of these errors. The plant-specific sensitivity study resulted in an estimated PCT impact of -10°F for Wolf Creek Generating Station (WCGS).
Attachment I to PA 15-0080 Page 2 of 2 WCGS CONTAINMENT COOLING CAPACITY Background Wolf Creek Nuclear Operating Corporation (WCNOC) identified an error in the containment fan cooler capacity transmitted for use in the best-estimate ASTRUM evaluation model analysis.This issue has been evaluated to estimate the impact on existing PCT results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.Affected Evaluation Models 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The estimated effect was determined for the large break LOCA evaluation model based on the change in calculated containment, pressure resulting from the correct containment cooling capacity.
The change in calculated containment pressure leads to an estimated effect of 0°F for the ASTRUM evaluation model analysis.
Attachment II to IRA 15-0080 Page 1 of 1 EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL PEAK CLADDING TEMPERATURE (PCT) MARGIN UTILIZATION RACK-UP FORM**LARGE BREAK LOCA PCT MARGIN UTILIZATION
- Evaluation Model: Fuel: Peaking Factor: SG Tube Plugging: Power Level: Limiting Break Size: LICENSING BASIS ASTRUM (2004)RFA-2 FQ=2.50, FdH=1.65 10%3565 MWth DEG Clad Temp (0 F)Ref. Notes Analysis of Record (AOR) PCT MARGIN ALLOCATIONS (APCT)1900°F 1 A. PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. None B. PLANNED PLANT CHANGE EVALUATIONS
- 1. Containment Fan Cooler Capacity C. 2014 PERMANENT ECCS MODEL ASSESSMENTS
- 1. Containment Fan Cooler Capacity 2. Decay Group Uncertainty Factors Errors D. OTHER 1. None 0 0 0-10 2 (a)2 3 0 LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1890 0 F
References:
- 1. WCAP-17107-P, Revision 1, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for the Wolf Creek Nuclear Power Plant Using the ASTRUM Methodology," January 2014.2. LTR-LIS-14-400, "10 CFR 50.46 Report for the Wolf Creek Large Break LOCA Evaluation of the Change in Containment Cooling Capacity," August 2014.3. LTR-LIS-14-492, "Wolf Creek Unit 1 10 CFR 50.46 Report for the Correction of the Decay Group Uncertainty Factors Errors," November 2014.Notes: (a) This effect was estimated based on a cooling capacity intended to bound future implementation of replacement tube bundles in the containment fan coolers.
