AEP-NRC-2016-17, Supplement to License Amendment Request to Adopt TSTF-425-A, Revision 3, Relocate Surveillance Frequencies to License Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5B.

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Supplement to License Amendment Request to Adopt TSTF-425-A, Revision 3, Relocate Surveillance Frequencies to License Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5B.
ML16039A240
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 02/04/2016
From: Lies Q
American Electric Power, Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2016-17
Download: ML16039A240 (12)
v • d • e


Text

INDIA NA Indiana Michigan Power Cook Nuclear Plant MICHIGAN One Cook Place POWER " Bridgman, MI 491fl6 AEP.com A unit of American ElectricPower February 4, 2016 AEP-NRC-201 6-17 10 CFR 50.4 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATT'N: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST TO ADOPT TSTF-425-A, REVISION 3, "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL - RISK INFORMED TECHNICAL SPECIFICATION TASK FORCE (RITSTF) INITIATIVE 5B"

Reference:

Letter from J. P. Gebbie, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant, Units 1 and 2 License Amendment Request to Adopt TSTF-425-A, Revision 3; 'Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative SB'," AEP-NRC-2015-46, dated November 19, 2015, Agencywide Documents Access and Management System Accession No. ML15328A450.

By the reference above, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, requested a license amendment to adopt Technical Specification Task Force (TSTF)-425-A, Revision 3, which relocates surveillance frequencies from technical specification to licensee control for CNP. During a telecon between A. W. Dietrich UIIS. Nuclear Regulatory Agency (NRC) and I&M staff, on December 11, 2015, the NRC requested that I&M supplement the application to address the Fire Probabilistic Risk Assessment (PRA) Facts and Observations (F&O) to complete the review of the reference above. to this letter provides an affirmation statement pertaining to the information contained herein. Enclosure 2 contains the supplemental information requested by the NRC. Enclosure 3 provides a table of new regulatory commitments.

U. S. Nuclear Regulatory Commission AEP-NRC-201 6-17 Page 2 Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Q!. hane Lies Site Vice President

'DMB/mll

Enclosures:

1. Affirmation 2.. Supplemental Information for the License Amendment Request to Adopt TSTF-425-A
3. Regulatory Commitments c: R. J. Ancona, MPSC A. W. Dietrich, NRC, Washington, D.C.

MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures

Enc-losure 1Ito AEP-NRC-2016-17 AFFI RMATION I, Q. Shane Lies, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of I&M, and that the statements made and forth herein pertaining to I&M are true and correct to the best of my knowledge, the matters set information, and belief.

Indiana Michigan Power Company Q.Shane Lies Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THI~S.

  • DAY OF1*£k ',>o,.Ž*'-,2016 Notary DANIELLE RURGOYNE Public, State of Mich 8,ar County of Eerrien "

My Commission Expires 0 4 - 4.20*8 Acting In the County oj .*** '¢ *.*

My Commission Expires Q)'\ - x' - 2. ,

Enclosure 2 to AEP-NRC-2016-17 Supplemental Information for the License Amendment Request to Adopt TSTF-425-.A

1.0 INTRODUCTION

Pursuant to 10 CFR 50.90, by letter dated November 19, 2015, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, submitted a license amendment request (LAR) to adopt Technical Specification Task Force (TSTF)-425-A, Revision 3 (Reference 3).

The U. S. Nuclear Regulatory Commission (NRC) staff reviewed the application and concluded that additional information is necessary to enable them to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment. During a telecon with A. W. Dietrich U. S. Nuclear Regulatory Agency (NRC) and I&M staff, on December 11, 2015, the NRC requested that I&M supplement the application to provide the new Fire Probabilistic Risk Assessment (PRA) Facts and Observations (F&Os) from a Focused Scope Peer Review.

report that I&M received after Reference 3 was submitted to the NRC.

2.0 OVERVIEW The full Technical Adequacy Justification of the Fire PRA was previously provided as part of the National Fire Protection Association 805 LAR (Reference 4). Attachment V, of Reference 4, provided the disposition of Peer Review F&Os from the Full Scope Peer Review performed on the Fire PRA in 2010.

Following the submittal of Reference 3, an additional Focused Scope Peer Review of the Fire PRA with respect to modeling of large early release frequency (LERF) was performed in November 2015, (Reference 1). The Focused Scope Peer Review assessed only the LERF (LE) and related Plant Response Model (PRM) requirements as related to the LERF portion of the Fire PRA model. As with the Internal Events PRA, each applicable supporting requirement (SR) in ASME RA-Sa-2009 (Reference 2) was evaluated against a goal of Capability Category (CC) I1. For each SR not meeting at least CC li, an evaluation is provided in the Fire PRA Focused Scope Peer Review Technical Adequacy Justification Table below with respect to its impact on the proposed Surveillance Frequency Control Program. The Focused Scope Peer Review is considered to supersede the 2010 Fire PRA Peer Review for graded SRs. As discussed in Reference 3, the July 2015, Full Scope Peer Review of the Internal Events PRA is considered to apply to the Fire PRA as well, in accordance with SR PRM-B2.

3.0 REFERENCES

1. ERIN Engineering and Research, Inc., "D. C. Cook Focused Scope Peer Review for Fire PRA,"' Document #D0403140002-1515, November 19, 2015.
2. ASME RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2, 2009.

to AEP-NRC-2016-17 Pg Page 2

3. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant, Units 1 and 2 License Amendment Request to Adopt TSTF-425-A, Revision 3, 'Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5B'," AEP-NRC-2015-46, dated November 19, 2015, Agencywide Documents Access and Management System (ADAMS) Accession No. ML15328A450.
4. Letter from M. H. Carlson, l&M to NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Docket Nos. 50-315 and 50-316, Request for License Amendment to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)," dated July 1, 2011, ADAMS No.

ML11188A145.

Fire Probabilistic Risk Assessment (PRA) Focused Scope Peer Review Technical Adequacy Justification Table Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-C1 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision 1. support applications for this SR.

LE-C2 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision 1 which is considered support applications conservative ratherthan realistic, for this SR.

LE-C3 INo repairof equipment after core damage CC I is considered to (Fire PRA) was considered. be sufficient to support applications for this SR.

__LE-~C4 IBasis: The CNP LERF analysis follows CC I is considered to (Fire PRA) methods in WCA P-I16341 and be sufficient to NUREG/CR-6595, Revision 1 and the support applications event trees developed in those reports. for this SR.

LE-C5 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision I which is considered support applications conservative ratherthan reallstic. for this SR.

to AEP-NRC-2016-17 Pg Page 3 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-C7 Not Met Sections 6. 5.3 and 6.5. 6 of The effects of fire on (Fire PRA) PRA-NB-FIRE-LE describes the operator LE-related operator actions credited in the LERF Model. The actions will be level of detail of the analysis does not reviewed and any reflect the use of the applicable necessary requirements in Section 2-2.5 of the PRA modifications will be Standard, added to the model prior to program implementation.

LE-C9 INo credit is taken for continued operation CC I is considered to (Fire PRA) of equipment or operatoractions in be sufficient to adverse environments, support applications for this SR.

LE-C1 0 INo credit is taken for survivability of CC I is considered to (Fire PRA) equipment or operatoractions in adverse be sufficient to environments, support applications for this SR.

LE-C11 IContainment failure equals LERF and CC I is considered to (Fire PRA) ends the analysis. No events beyond be sufficient to containment failure are postulated, support applications for this SR.

LE-C12 IContainment failure equals LERF and CC I is considered to (Fire PRA) ends the analysis. No continued be sufficient to operation of equipment beyond support applications containment failure is postulated. for this SR.

LE-C13 IBypass was a deterministic event (YES or CC I is considered to (Fire PRA) NO). No source terms or scrubbing or be sufficient to decontaminationwas evaluated. All support applications Steam Generator Tube Rupture for this SR.

sequences go to LERF.

to AEP-NRC-2016-17 Pg Page 4 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-D1 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment I of PRA-L2 probabilities given MODEL, Revision 0) that predicts the hydrogen igniter ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clearif and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not discussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appears to be using containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications for this SR.

LE-D2 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment 1 of PRA-L2 probabilities given MODEL, Revision 0) that predicts the hydrogen igniter ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clear if and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not dis~cussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appears to be using -containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications for this SR.

to AEP-NRC-2016-17 Pg Page 5 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal.

Requirement Category Assessment (SR) (CC)

LE-D3 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment 1 of probabilities given PRA-L2 MODEL, Revision 0) that predicts hydrogen igniter the ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clear if and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not discussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appearsto be using containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications

" _ .for this SR.

LE-D5 IModels from WCAP-16341 are used for CC I is considered to (Fire PRA) TI-SGTR and PI-SGTR. SGTR initiator be sufficient to taken directly to containment bypass. support applications for this SR.

Secondary Side Isolation is not considered to result in a direct containment bypass.

LE-EI Not Met Sources for parametervalues are shown The effects of fire on (Fire PRA) in Table 2 of PRA-NB-FIRE-LE. LE-related operator Appropriate parametervalues were actions will be selected consistent with the requirements reviewed and any of technical element DA. Operatoractions necessary identified in Sections _6.5.3 an~d 6.5.6 were modifications will be not selected in accordance with Section added *t6 the ;model 2-2.5 of the PRA Standard. prior to program implementation.

LE-E2 IBasis: Data is taken from CC I is considered to (Fire PRA) NUREG/CR-6595 or WCAP-16341. be sufficient to support applications for this SR.

to AEP-NRC-2016-17 Pg Page 6 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-F1 Not Met The results of the LERF quantification and Resolution of this SR (Fire PRA) cutset reviews are provided in will provide additional PRA -FIRE-I17663-014-MAR-RI-finalI-1017, analysis of results, but Tables 5-I, 5-7; 5-9, 5-17 and 5-19. The will not impact the results do not provide contributionsby actual results.

LERF PBS designation and LERF failure Therefore, the mechanism. improvements will be documentation improvements and will not impact the use of this application.

LE-F2 Not Met The CNP results were not compared to a Resolution of this SR (Fire PRA) peer plant. is expected to involve documentation improvements only since the comparison was already performed at a high level. Containment failure probabilities given hydrogen igniter failure are taken from NUREG/CR-6595, and no issues were noted with the hydrogen igniter system model.

LE-G3 Not Met The results of the LERF quantification and Resolution of this SR (Fire PRA) cutset reviews are provided in will provide additional PRA -FIRE-I17663-014-LAR-RI-finaI-1017, analysis of results, but Tables 5-1, 5-7, 5-9, 5-17 and 5-19. The will not impact the results do not provide contributions by actual results.

LERF plant damage state designation Therefore, the and LERF failure mechanism. improvements will be documentation improvements and will not impact the use of this application.

to AEP-NRC-2016-17 Pg Page 7 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-G6 Not Met Sections 5.2 and 5.3 of Resolution of this SR (Fire PRA) PRA-FIRE-1 7663-O14-LAR-RI-finaI-10l 7 is expected to involve provide a quantitative definition used for documentation significant core damage accident improvements only progressionsequence that is consistent and therefore will not with Part 1-2 of the standard.However, impact the use of this there is no equivalent definition for LERF. application.

PRM-B2 Not Met An assessment of Internal Event PRA As full power internal (Fire PRA) peer review deficiencies is requiredto events F&Os are evaluate the impact on the Fire PRA. resolved, their impact on the Fire PRA will also be evaluated.

Reintegration of the Fire PRA (Reference

3) will resolve the relevant F&Os in the Fire PRA PRM-B14 Not Met Provide documentation demonstratingan The effects of fire on (Fire PRA) evaluation for this SR. Evaluate the LERF bypass potential for screened LERF scenarios pathways will be impacting the Fire PRA, e.g., LERF reviewed and any bypass pathway screened based on size, necessary where a fire may impact multiple modifications will be pathways where the sum of the pathway added to the model szsmay exceed the LERF bypass iplementtionpoga pathway screening criteria. mlmnain to AEP-NRC-2016-17 Pg Page 8 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category IAssessment (SR) (CC) _______________j________

PRM-B1 5 Not Met Provide documentation demonstratingan This SR requires (Fire PRA) evaluation for this SR. documentation that the systems analysis, accident sequence analysis, and human reliability analysis in the Fire LERF model meets the relevant requirements of Part 2 of ASME/ANS-RA-Sa-2009 in the context of fire events. The Fire LERF notebook will be updated to include documentation of these requirements.

prior to program implementation.

Enclosure 3 to AEP-NRC-2016-17 REGULATORY COMMITMENTS The following table identifies an action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. TheY are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments. All commitments discussed in this table are one-time commitments.

Commitment Scheduled Completion Date (if applicable)

Implement the resolution for the following Supporting Requirements Prior to program from Enclosure 2, Fire Probabilistic Risk Assessment (PRA) Focused implementation Scope Peer Review Technical Adequacy Justification Table:

LE-C7, LE-EI, PRM-B14, and PRM-B15 _________

INDIA NA Indiana Michigan Power Cook Nuclear Plant MICHIGAN One Cook Place POWER " Bridgman, MI 491fl6 AEP.com A unit of American ElectricPower February 4, 2016 AEP-NRC-201 6-17 10 CFR 50.4 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATT'N: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 SUPPLEMENT TO LICENSE AMENDMENT REQUEST TO ADOPT TSTF-425-A, REVISION 3, "RELOCATE SURVEILLANCE FREQUENCIES TO LICENSEE CONTROL - RISK INFORMED TECHNICAL SPECIFICATION TASK FORCE (RITSTF) INITIATIVE 5B"

Reference:

Letter from J. P. Gebbie, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant, Units 1 and 2 License Amendment Request to Adopt TSTF-425-A, Revision 3; 'Relocate Surveillance Frequencies to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative SB'," AEP-NRC-2015-46, dated November 19, 2015, Agencywide Documents Access and Management System Accession No. ML15328A450.

By the reference above, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, requested a license amendment to adopt Technical Specification Task Force (TSTF)-425-A, Revision 3, which relocates surveillance frequencies from technical specification to licensee control for CNP. During a telecon between A. W. Dietrich UIIS. Nuclear Regulatory Agency (NRC) and I&M staff, on December 11, 2015, the NRC requested that I&M supplement the application to address the Fire Probabilistic Risk Assessment (PRA) Facts and Observations (F&O) to complete the review of the reference above. to this letter provides an affirmation statement pertaining to the information contained herein. Enclosure 2 contains the supplemental information requested by the NRC. Enclosure 3 provides a table of new regulatory commitments.

U. S. Nuclear Regulatory Commission AEP-NRC-201 6-17 Page 2 Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Q!. hane Lies Site Vice President

'DMB/mll

Enclosures:

1. Affirmation 2.. Supplemental Information for the License Amendment Request to Adopt TSTF-425-A
3. Regulatory Commitments c: R. J. Ancona, MPSC A. W. Dietrich, NRC, Washington, D.C.

MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures

Enc-losure 1Ito AEP-NRC-2016-17 AFFI RMATION I, Q. Shane Lies, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of I&M, and that the statements made and forth herein pertaining to I&M are true and correct to the best of my knowledge, the matters set information, and belief.

Indiana Michigan Power Company Q.Shane Lies Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THI~S.

  • DAY OF1*£k ',>o,.Ž*'-,2016 Notary DANIELLE RURGOYNE Public, State of Mich 8,ar County of Eerrien "

My Commission Expires 0 4 - 4.20*8 Acting In the County oj .*** '¢ *.*

My Commission Expires Q)'\ - x' - 2. ,

Enclosure 2 to AEP-NRC-2016-17 Supplemental Information for the License Amendment Request to Adopt TSTF-425-.A

1.0 INTRODUCTION

Pursuant to 10 CFR 50.90, by letter dated November 19, 2015, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, submitted a license amendment request (LAR) to adopt Technical Specification Task Force (TSTF)-425-A, Revision 3 (Reference 3).

The U. S. Nuclear Regulatory Commission (NRC) staff reviewed the application and concluded that additional information is necessary to enable them to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment. During a telecon with A. W. Dietrich U. S. Nuclear Regulatory Agency (NRC) and I&M staff, on December 11, 2015, the NRC requested that I&M supplement the application to provide the new Fire Probabilistic Risk Assessment (PRA) Facts and Observations (F&Os) from a Focused Scope Peer Review.

report that I&M received after Reference 3 was submitted to the NRC.

2.0 OVERVIEW The full Technical Adequacy Justification of the Fire PRA was previously provided as part of the National Fire Protection Association 805 LAR (Reference 4). Attachment V, of Reference 4, provided the disposition of Peer Review F&Os from the Full Scope Peer Review performed on the Fire PRA in 2010.

Following the submittal of Reference 3, an additional Focused Scope Peer Review of the Fire PRA with respect to modeling of large early release frequency (LERF) was performed in November 2015, (Reference 1). The Focused Scope Peer Review assessed only the LERF (LE) and related Plant Response Model (PRM) requirements as related to the LERF portion of the Fire PRA model. As with the Internal Events PRA, each applicable supporting requirement (SR) in ASME RA-Sa-2009 (Reference 2) was evaluated against a goal of Capability Category (CC) I1. For each SR not meeting at least CC li, an evaluation is provided in the Fire PRA Focused Scope Peer Review Technical Adequacy Justification Table below with respect to its impact on the proposed Surveillance Frequency Control Program. The Focused Scope Peer Review is considered to supersede the 2010 Fire PRA Peer Review for graded SRs. As discussed in Reference 3, the July 2015, Full Scope Peer Review of the Internal Events PRA is considered to apply to the Fire PRA as well, in accordance with SR PRM-B2.

3.0 REFERENCES

1. ERIN Engineering and Research, Inc., "D. C. Cook Focused Scope Peer Review for Fire PRA,"' Document #D0403140002-1515, November 19, 2015.
2. ASME RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, February 2, 2009.

to AEP-NRC-2016-17 Pg Page 2

3. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant, Units 1 and 2 License Amendment Request to Adopt TSTF-425-A, Revision 3, 'Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5B'," AEP-NRC-2015-46, dated November 19, 2015, Agencywide Documents Access and Management System (ADAMS) Accession No. ML15328A450.
4. Letter from M. H. Carlson, l&M to NRC, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Docket Nos. 50-315 and 50-316, Request for License Amendment to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)," dated July 1, 2011, ADAMS No.

ML11188A145.

Fire Probabilistic Risk Assessment (PRA) Focused Scope Peer Review Technical Adequacy Justification Table Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-C1 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision 1. support applications for this SR.

LE-C2 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision 1 which is considered support applications conservative ratherthan realistic, for this SR.

LE-C3 INo repairof equipment after core damage CC I is considered to (Fire PRA) was considered. be sufficient to support applications for this SR.

__LE-~C4 IBasis: The CNP LERF analysis follows CC I is considered to (Fire PRA) methods in WCA P-I16341 and be sufficient to NUREG/CR-6595, Revision 1 and the support applications event trees developed in those reports. for this SR.

LE-C5 IThe CNP LERF analysis follows methods CC I is considered to (Fire PRA) in WCAP-16341 and NUREG/CR-6595, be sufficient to Revision I which is considered support applications conservative ratherthan reallstic. for this SR.

to AEP-NRC-2016-17 Pg Page 3 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-C7 Not Met Sections 6. 5.3 and 6.5. 6 of The effects of fire on (Fire PRA) PRA-NB-FIRE-LE describes the operator LE-related operator actions credited in the LERF Model. The actions will be level of detail of the analysis does not reviewed and any reflect the use of the applicable necessary requirements in Section 2-2.5 of the PRA modifications will be Standard, added to the model prior to program implementation.

LE-C9 INo credit is taken for continued operation CC I is considered to (Fire PRA) of equipment or operatoractions in be sufficient to adverse environments, support applications for this SR.

LE-C1 0 INo credit is taken for survivability of CC I is considered to (Fire PRA) equipment or operatoractions in adverse be sufficient to environments, support applications for this SR.

LE-C11 IContainment failure equals LERF and CC I is considered to (Fire PRA) ends the analysis. No events beyond be sufficient to containment failure are postulated, support applications for this SR.

LE-C12 IContainment failure equals LERF and CC I is considered to (Fire PRA) ends the analysis. No continued be sufficient to operation of equipment beyond support applications containment failure is postulated. for this SR.

LE-C13 IBypass was a deterministic event (YES or CC I is considered to (Fire PRA) NO). No source terms or scrubbing or be sufficient to decontaminationwas evaluated. All support applications Steam Generator Tube Rupture for this SR.

sequences go to LERF.

to AEP-NRC-2016-17 Pg Page 4 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-D1 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment I of PRA-L2 probabilities given MODEL, Revision 0) that predicts the hydrogen igniter ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clearif and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not discussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appears to be using containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications for this SR.

LE-D2 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment 1 of PRA-L2 probabilities given MODEL, Revision 0) that predicts the hydrogen igniter ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clear if and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not dis~cussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appears to be using -containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications for this SR.

to AEP-NRC-2016-17 Pg Page 5 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal.

Requirement Category Assessment (SR) (CC)

LE-D3 ICNP has a plant-specific containment Containment failure (Fire PRA) fragility analysis (Attachment 1 of probabilities given PRA-L2 MODEL, Revision 0) that predicts hydrogen igniter the ultimate containment capacity and the failure are taken from location of containment failure on NUREG/CR-6595, pressure. However, it is not clear if and and no issues were how this calculation was factored into the noted with the simplified Level 2 model documented in hydrogen igniter PRA-NB-FIRE-LE. Attachment 1 of system model. This PRA-L2 MODEL is not cited in Section 8 aspect produces a of PRA-NB-FIRE-LE and not discussed in possibly conservative Section 6.5.1 of the report. The simplified estimate of Level 2 model appearsto be using containment failure, NUREG/CR-6595 if the igniters fail. so CC I is considered to be sufficient to support applications

" _ .for this SR.

LE-D5 IModels from WCAP-16341 are used for CC I is considered to (Fire PRA) TI-SGTR and PI-SGTR. SGTR initiator be sufficient to taken directly to containment bypass. support applications for this SR.

Secondary Side Isolation is not considered to result in a direct containment bypass.

LE-EI Not Met Sources for parametervalues are shown The effects of fire on (Fire PRA) in Table 2 of PRA-NB-FIRE-LE. LE-related operator Appropriate parametervalues were actions will be selected consistent with the requirements reviewed and any of technical element DA. Operatoractions necessary identified in Sections _6.5.3 an~d 6.5.6 were modifications will be not selected in accordance with Section added *t6 the ;model 2-2.5 of the PRA Standard. prior to program implementation.

LE-E2 IBasis: Data is taken from CC I is considered to (Fire PRA) NUREG/CR-6595 or WCAP-16341. be sufficient to support applications for this SR.

to AEP-NRC-2016-17 Pg Page 6 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category Assessment (SR) (CC)

LE-F1 Not Met The results of the LERF quantification and Resolution of this SR (Fire PRA) cutset reviews are provided in will provide additional PRA -FIRE-I17663-014-MAR-RI-finalI-1017, analysis of results, but Tables 5-I, 5-7; 5-9, 5-17 and 5-19. The will not impact the results do not provide contributionsby actual results.

LERF PBS designation and LERF failure Therefore, the mechanism. improvements will be documentation improvements and will not impact the use of this application.

LE-F2 Not Met The CNP results were not compared to a Resolution of this SR (Fire PRA) peer plant. is expected to involve documentation improvements only since the comparison was already performed at a high level. Containment failure probabilities given hydrogen igniter failure are taken from NUREG/CR-6595, and no issues were noted with the hydrogen igniter system model.

LE-G3 Not Met The results of the LERF quantification and Resolution of this SR (Fire PRA) cutset reviews are provided in will provide additional PRA -FIRE-I17663-014-LAR-RI-finaI-1017, analysis of results, but Tables 5-1, 5-7, 5-9, 5-17 and 5-19. The will not impact the results do not provide contributions by actual results.

LERF plant damage state designation Therefore, the and LERF failure mechanism. improvements will be documentation improvements and will not impact the use of this application.

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LE-G6 Not Met Sections 5.2 and 5.3 of Resolution of this SR (Fire PRA) PRA-FIRE-1 7663-O14-LAR-RI-finaI-10l 7 is expected to involve provide a quantitative definition used for documentation significant core damage accident improvements only progressionsequence that is consistent and therefore will not with Part 1-2 of the standard.However, impact the use of this there is no equivalent definition for LERF. application.

PRM-B2 Not Met An assessment of Internal Event PRA As full power internal (Fire PRA) peer review deficiencies is requiredto events F&Os are evaluate the impact on the Fire PRA. resolved, their impact on the Fire PRA will also be evaluated.

Reintegration of the Fire PRA (Reference

3) will resolve the relevant F&Os in the Fire PRA PRM-B14 Not Met Provide documentation demonstratingan The effects of fire on (Fire PRA) evaluation for this SR. Evaluate the LERF bypass potential for screened LERF scenarios pathways will be impacting the Fire PRA, e.g., LERF reviewed and any bypass pathway screened based on size, necessary where a fire may impact multiple modifications will be pathways where the sum of the pathway added to the model szsmay exceed the LERF bypass iplementtionpoga pathway screening criteria. mlmnain to AEP-NRC-2016-17 Pg Page 8 Supporting Capability Peer Review Assessment Basis TSTF-425 Submittal Requirement Category IAssessment (SR) (CC) _______________j________

PRM-B1 5 Not Met Provide documentation demonstratingan This SR requires (Fire PRA) evaluation for this SR. documentation that the systems analysis, accident sequence analysis, and human reliability analysis in the Fire LERF model meets the relevant requirements of Part 2 of ASME/ANS-RA-Sa-2009 in the context of fire events. The Fire LERF notebook will be updated to include documentation of these requirements.

prior to program implementation.

Enclosure 3 to AEP-NRC-2016-17 REGULATORY COMMITMENTS The following table identifies an action committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. TheY are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments. All commitments discussed in this table are one-time commitments.

Commitment Scheduled Completion Date (if applicable)

Implement the resolution for the following Supporting Requirements Prior to program from Enclosure 2, Fire Probabilistic Risk Assessment (PRA) Focused implementation Scope Peer Review Technical Adequacy Justification Table:

LE-C7, LE-EI, PRM-B14, and PRM-B15 _________

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