ML12345A327: Difference between revisions

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| document type = Request for Additional Information (RAI), Letter
| document type = Request for Additional Information (RAI), Letter
| page count = 5
| page count = 5
| project = TAC:ME6630, TAC:ME6629
| project = TAC:ME6629, TAC:ME6630
| stage = RAI
| stage = RAI
}}
}}
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==Dear Mr. Weber:==
==Dear Mr. Weber:==
By letter dated July 1, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 11188A145), as supplemented by letters dated September 2,2011, April 27, 2012, June 29,2012, August 9,2012, October 15,2012, and November 9,2012, Indiana Michigan Power Company (I&M) submitted an application for a license amendment to transition the Donald C. Cook Nuclear Plant, Units 1 and 2, fire protection program, from Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(b), to 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805. Some portions of the supplemental information contain security related information and are therefore withheld from public disclosure. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject submittals and determined that additional information is needed to complete the review, as described in the enclosure, request for additional information (RAI). The NRC staff had discussed the RAI in draft form with your staff on December 6, 2012. During that discussion, the NRC staff agreed to revise certain draft questions. Your staff also agreed to formally submit your response within 30 days from the date of this letter. Please note that the NRC staff's review efforts are continuing and additional RAls may be forthcoming.
By letter dated July 1, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 11188A145), as supplemented by letters dated September 2,2011, April 27, 2012, June 29,2012, August 9,2012, October 15,2012, and November 9,2012, Indiana Michigan Power Company (I&M) submitted an application for a license amendment to transition the Donald C. Cook Nuclear Plant, Units 1 and 2, fire protection program, from Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(b), to 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805. Some portions of the supplemental information contain security related information and are therefore withheld from public disclosure.
L. Weber -Please feel free to contact me if you need any further clarification of the questions in the enclosure. Sincerely. Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ REQUEST FOR ADDITIONAL INFORMATION (RAil LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316 RAI 20.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 and the enclosure, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-20, Probabilistic Risk Assessment (PRA), Internal Events PRA (lEPRA), explaining that a gap assessment was performed on the differences in supporting requirements (SRs) between PRA Standard RA-Sa-2009, as clarified by Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Result for Risk-Informed Activities", Rev, 2, and RA-Sa-2003, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (which was initially used to review the internal events PRA). As part of the assessment, gap findings were generated and a sensitivity study was performed to address a subset of the issues. Clarify the following gap finding: A finding against SR LE-B3, which requires using engineering analysis to support PRA modeling, identifies two large early release frequency (LERF) modeling assumptions that were not justified with engineering assessment but have beneficial impacts on the LERF estimate. The dispOSition for this gap finding states that these assumptions are expected to be confirmed by analysis. The assumptions are not identified. Identify the assumptions made, and explain why they are expected to be confirmed by analysis. A finding against SR LE-A4, which requires dependencies between the Level 1 and Level 2 PRA to be properly treated, identifies that in some cases, assumptions were made that may yield non-conservative results. The disposition to this finding states that the licensee's use of NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events," produces a bounding model but also states that a sensitivity study examines the impact of this issue. The finding against LE-A4 is not identified in the sensitivity study identified for this RAI on page 22 of the response. Clarify how the finding against LE-A4 is being dispositioned. Include in this description how dependencies between Level 1 and Level 2 are treated and clarify how SR LE-A4 is met. Enclosure   
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject submittals and determined that additional information is needed to complete the review, as described in the enclosure, request for additional information (RAI). The NRC staff had discussed the RAI in draft form with your staff on December 6, 2012. During that discussion, the NRC staff agreed to revise certain draft questions.
-RAI 30.01 Probabilistic Risk Assessment In your letter dated April 27, 2012 (ADAMS Accession No. ML 12132A390) you responded to RAI-30, Probabilistic Risk Assessment, Uncertainty, discussing the results of sensitivity analyses using frequently asked question (FAQ) 08-0048 (ADAMS Accession No. ML092190457) fire ignition frequencies and the substantial risk reduction obtained from crediting a local operator action to restore the turbine driven auxiliary feedwater pump N train battery charger after it is tripped following loss of offsite power or after a safety injection signal. Given that including this action reduces the sensitivity study results to below the risk acceptance criteria in RG 1.174, discuss whether the RA is or will be included in the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," fire protection program. If included, provide the additional risk of this RA. If not included in the NFPA 805 fire protection program, provide justification for not including it. RAI 34.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 (ADAMS Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-34e, Probabilistic Risk Assessment, Peer Review F&Os, discussing the results of sensitivity analyses that shows that using the "Special Weighting Factors" for transient fire frequency apportionment compared to using the nominal 6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities," (no fractional values) method will result in the delta core damage frequency (CDF) exceeding the RG 1.174 risk acceptance criteria. Provide the corresponding results for LERF and delta LERF.
Your staff also agreed to formally submit your response within 30 days from the date of this letter. Please note that the NRC staff's review efforts are continuing and additional RAls may be forthcoming.
L. Weber -2 Please feel free to contact me if you need any further clarification of the questions in the enclosure. Sincerely, IRA! Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ DISTRIBUTION: PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 Resource RidsNrrPMDCCook Resource RidsNrrLABTully Resource RidsOgcRp Resource RidsAcrsAcnw _MaiICTR Resource RidsNrrDraAfpb Resource RidsNrrDorlDpr Resource RidsRgn3MailCenter Resource RidsNrrDraApla Resource MSnodderly, NRR DHarrison, NRR AKlein, NRR LFields, NRR SDinsmore, NRR HBarrett, NRR JRobinson, NRR . db)y J R b' Inson e-mal 0 ADAMS A ccesslon N0.: ML 12345A327 '"RAI transmltte , o 'I f11/27/12 OFFICE LPL3-1/PM LPL3-1LAit LPL3-21lA DRAlAPLAlBC LPL3-1/BC LPL3-1/PM NAME lWenQert ELee SRohrer DHarrison* RCarlson MChaw/a for lWenQert DATE 12/13/12 12/13/12 12/13/12 11/27/12 12/13/12 12/13/12 OFFICIAL RECORD
L. Weber -Please feel free to contact me if you need any further clarification of the questions in the enclosure.
}}
Sincerely.
Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ REQUEST FOR ADDITIONAL INFORMATION (RAil LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316 RAI 20.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 and the enclosure, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-20, Probabilistic Risk Assessment (PRA), Internal Events PRA (lEPRA), explaining that a gap assessment was performed on the differences in supporting requirements (SRs) between PRA Standard RA-Sa-2009, as clarified by Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Result for Risk-Informed Activities", Rev, 2, and RA-Sa-2003, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (which was initially used to review the internal events PRA). As part of the assessment, gap findings were generated and a sensitivity study was performed to address a subset of the issues. Clarify the following gap finding: A finding against SR LE-B3, which requires using engineering analysis to support PRA modeling, identifies two large early release frequency (LERF) modeling assumptions that were not justified with engineering assessment but have beneficial impacts on the LERF estimate.
The dispOSition for this gap finding states that these assumptions are expected to be confirmed by analysis.
The assumptions are not identified.
Identify the assumptions made, and explain why they are expected to be confirmed by analysis. A finding against SR LE-A4, which requires dependencies between the Level 1 and Level 2 PRA to be properly treated, identifies that in some cases, assumptions were made that may yield non-conservative results. The disposition to this finding states that the licensee's use of NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events," produces a bounding model but also states that a sensitivity study examines the impact of this issue. The finding against LE-A4 is not identified in the sensitivity study identified for this RAI on page 22 of the response.
Clarify how the finding against LE-A4 is being dispositioned.
Include in this description how dependencies between Level 1 and Level 2 are treated and clarify how SR LE-A4 is met. Enclosure   
-RAI 30.01 Probabilistic Risk Assessment In your letter dated April 27, 2012 (ADAMS Accession No. ML 12132A390) you responded to RAI-30, Probabilistic Risk Assessment, Uncertainty, discussing the results of sensitivity analyses using frequently asked question (FAQ) 08-0048 (ADAMS Accession No. ML092190457) fire ignition frequencies and the substantial risk reduction obtained from crediting a local operator action to restore the turbine driven auxiliary feedwater pump N train battery charger after it is tripped following loss of offsite power or after a safety injection signal. Given that including this action reduces the sensitivity study results to below the risk acceptance criteria in RG 1.174, discuss whether the RA is or will be included in the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," fire protection program. If included, provide the additional risk of this RA. If not included in the NFPA 805 fire protection program, provide justification for not including it. RAI 34.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 (ADAMS Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-34e, Probabilistic Risk Assessment, Peer Review F&Os, discussing the results of sensitivity analyses that shows that using the "Special Weighting Factors" for transient fire frequency apportionment compared to using the nominal 6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities," (no fractional values) method will result in the delta core damage frequency (CDF) exceeding the RG 1.174 risk acceptance criteria.
Provide the corresponding results for LERF and delta LERF.
L. Weber -2 Please feel free to contact me if you need any further clarification of the questions in the enclosure.
Sincerely, IRA! Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 Resource RidsNrrPMDCCook Resource RidsNrrLABTully Resource RidsOgcRp Resource RidsAcrsAcnw
_MaiICTR Resource RidsNrrDraAfpb Resource RidsNrrDorlDpr Resource RidsRgn3MailCenter Resource RidsNrrDraApla Resource MSnodderly, NRR DHarrison, NRR AKlein, NRR LFields, NRR SDinsmore, NRR HBarrett, NRR JRobinson, NRR . db)y J R b' Inson e-mal 0 ADAMS A ccesslon N 0.: ML 12345A327  
'"RAI t ransmltte , o 'I f11/27/12 OFFICE LPL3-1/PM LPL3-1LAit LPL3-21lA DRAlAPLAlBC LPL3-1/BC LPL3-1/PM NAME lWenQert ELee SRohrer DHarrison*
RCarlson MChaw/a for lWenQert DATE 12/13/12 12/13/12 12/13/12 11/27/12 12/13/12 12/13/12 OFFICIAL RECORD}}

Revision as of 06:00, 22 July 2018

D. C. Cook Nuclear Plant, Units 1 and 2, Request for Additional Information on the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805
ML12345A327
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 12/13/2012
From: Wengert T J
Plant Licensing Branch III
To: Weber L J
Nuclear Generation Group
Wengert T J
References
TAC ME6629, TAC ME6630
Download: ML12345A327 (5)
v • d • e


Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 December 13, 2012 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 -REQUEST FOR ADDITIONAL INFORMATION ON THE APPLICATION FOR AMENDMENT TO TRANSITION THE FIRE PROTECTION PROGRAM TO NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 (TAC NOS. ME6629 AND ME6630)

Dear Mr. Weber:

By letter dated July 1, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 11188A145), as supplemented by letters dated September 2,2011, April 27, 2012, June 29,2012, August 9,2012, October 15,2012, and November 9,2012, Indiana Michigan Power Company (I&M) submitted an application for a license amendment to transition the Donald C. Cook Nuclear Plant, Units 1 and 2, fire protection program, from Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48(b), to 10 CFR 50.48(c), National Fire Protection Association Standard (NFPA) 805. Some portions of the supplemental information contain security related information and are therefore withheld from public disclosure.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject submittals and determined that additional information is needed to complete the review, as described in the enclosure, request for additional information (RAI). The NRC staff had discussed the RAI in draft form with your staff on December 6, 2012. During that discussion, the NRC staff agreed to revise certain draft questions.

Your staff also agreed to formally submit your response within 30 days from the date of this letter. Please note that the NRC staff's review efforts are continuing and additional RAls may be forthcoming.

L. Weber -Please feel free to contact me if you need any further clarification of the questions in the enclosure.

Sincerely.

Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ REQUEST FOR ADDITIONAL INFORMATION (RAil LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316 RAI 20.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 and the enclosure, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-20, Probabilistic Risk Assessment (PRA), Internal Events PRA (lEPRA), explaining that a gap assessment was performed on the differences in supporting requirements (SRs) between PRA Standard RA-Sa-2009, as clarified by Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Result for Risk-Informed Activities", Rev, 2, and RA-Sa-2003, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (which was initially used to review the internal events PRA). As part of the assessment, gap findings were generated and a sensitivity study was performed to address a subset of the issues. Clarify the following gap finding: A finding against SR LE-B3, which requires using engineering analysis to support PRA modeling, identifies two large early release frequency (LERF) modeling assumptions that were not justified with engineering assessment but have beneficial impacts on the LERF estimate.

The dispOSition for this gap finding states that these assumptions are expected to be confirmed by analysis.

The assumptions are not identified.

Identify the assumptions made, and explain why they are expected to be confirmed by analysis. A finding against SR LE-A4, which requires dependencies between the Level 1 and Level 2 PRA to be properly treated, identifies that in some cases, assumptions were made that may yield non-conservative results. The disposition to this finding states that the licensee's use of NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events," produces a bounding model but also states that a sensitivity study examines the impact of this issue. The finding against LE-A4 is not identified in the sensitivity study identified for this RAI on page 22 of the response.

Clarify how the finding against LE-A4 is being dispositioned.

Include in this description how dependencies between Level 1 and Level 2 are treated and clarify how SR LE-A4 is met. Enclosure

-RAI 30.01 Probabilistic Risk Assessment In your letter dated April 27, 2012 (ADAMS Accession No. ML 12132A390) you responded to RAI-30, Probabilistic Risk Assessment, Uncertainty, discussing the results of sensitivity analyses using frequently asked question (FAQ) 08-0048 (ADAMS Accession No. ML092190457) fire ignition frequencies and the substantial risk reduction obtained from crediting a local operator action to restore the turbine driven auxiliary feedwater pump N train battery charger after it is tripped following loss of offsite power or after a safety injection signal. Given that including this action reduces the sensitivity study results to below the risk acceptance criteria in RG 1.174, discuss whether the RA is or will be included in the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," fire protection program. If included, provide the additional risk of this RA. If not included in the NFPA 805 fire protection program, provide justification for not including it. RAI 34.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 (ADAMS Accession No. ML 12242A246) and the non-public enclosure, you responded to RAI-34e, Probabilistic Risk Assessment, Peer Review F&Os, discussing the results of sensitivity analyses that shows that using the "Special Weighting Factors" for transient fire frequency apportionment compared to using the nominal 6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities," (no fractional values) method will result in the delta core damage frequency (CDF) exceeding the RG 1.174 risk acceptance criteria.

Provide the corresponding results for LERF and delta LERF.

L. Weber -2 Please feel free to contact me if you need any further clarification of the questions in the enclosure.

Sincerely, IRA! Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and Request for Additional Information cc: Distribution via ListServ DISTRIBUTION:

PUBLIC LPL3-1 r/f RidsNrrDorlLpl3-1 Resource RidsNrrPMDCCook Resource RidsNrrLABTully Resource RidsOgcRp Resource RidsAcrsAcnw

_MaiICTR Resource RidsNrrDraAfpb Resource RidsNrrDorlDpr Resource RidsRgn3MailCenter Resource RidsNrrDraApla Resource MSnodderly, NRR DHarrison, NRR AKlein, NRR LFields, NRR SDinsmore, NRR HBarrett, NRR JRobinson, NRR . db)y J R b' Inson e-mal 0 ADAMS A ccesslon N 0.: ML 12345A327

'"RAI t ransmltte , o 'I f11/27/12 OFFICE LPL3-1/PM LPL3-1LAit LPL3-21lA DRAlAPLAlBC LPL3-1/BC LPL3-1/PM NAME lWenQert ELee SRohrer DHarrison*

RCarlson MChaw/a for lWenQert DATE 12/13/12 12/13/12 12/13/12 11/27/12 12/13/12 12/13/12 OFFICIAL RECORD

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