AEP-NRC-2013-17, Response to a Request for Additional Information (RAI 29.01, 61, and 62) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805

Response to a Request for Additional Information (RAI 29.01, 61, and 62) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805

ML13123A298
Person / Time
Site:	Cook
Issue date:	05/01/2013
From:	Gebbie J Indiana & Michigan Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2013-17, TAC ME6629, TAC ME6630
Download: ML13123A298 (41)

Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWfER One Cook Place A unit of American Electric Power Bridgman, MI 49106 IndianaMichiganPower.com May 1, 2013 AEP-NRC-2013-17 10 CFR 50.90 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 Response to a Request for Additional Information (RAI 29.01, 61, and 62) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805 (TAC Nos. ME6629 and ME6630)

This letter provides Indiana Michigan Power Company's (l&M's), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to Requests for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding a proposed license amendment to transition CNP, Units 1 and 2, to a new fire protection program based on National Fire Protection Association Standard 805 (NFPA 805). Documents referenced in this letter and its enclosures are identified in.

By References 1 and 2, I&M proposed to amend CNP Units 1 and 2 Facility Operating Licenses DPR-58 and DPR-74 to adopt a new fire protection program based on NFPA 805, in accordance with 10 CFR 50.48(a) and (c). Reference 1, hereafter referred to as the Transition Report, provided information associated with the CNP transition to NFPA 805. By References 3, 4, 8, 9, 11, and 14, the NRC transmitted RAIs regarding the proposed amendment. References 5, 6, 7, 12, 13, 15, and 16 transmitted I&M's responses to the Reference 3, 4, 11, and 14 RAIs. This letter provides I&M's response to the Reference 8 RAI-29.01 and 61, Reference 9 RAI-62. Additionally, it provides I&M's NFPA 805 License Amendments Request (LAR) Revisions. to this letter provides an affirmation statement. identifies documents referenced in this letter and its enclosures. Enclosure 3 provides I&M's response to RAI-29.01. provides I&M's response to RAI-61. Enclosure 5 provides I&M's response to RAI-62. provides I&M's NFPA 805 LAR Revisions. provides a revised CNP Transition Report, Section 5.4, Transition Implementation Schedule, that reflects a twelve-month transition period. Enclosures 8 and 9 provide revisions to Attachments M and S, respectively, of the Transition Report in support of RAI's 29.01, 61, and 62, and the revised twelve month implementation schedule. Enclosures 10 and 11 provide I&M's revised portions of Attachments C Security-Related Information -Withhold From Public Disclosure Under 10 CFR 2.390.

Enclosures 10 and 11 to this letter contain security-related information.

Upon removal of Enclosures 10 and 11, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission AEP-NRC-2013-17 Page 2 and G of the Transition Report that reflect corrections of inconsistencies with support documentation.

I&M requests that Enclosures 10 and 11, which contain sensitive security related information, be withheld from public disclosure in accordance with 10 CFR 2.390.

Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

There are four new License Conditions and no new regulatory commitments associated with this response.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President DMB/kmh

Enclosures:

1. Affirmation

2. Identification of Documents Referenced in this Letter and its Enclosures

3. Response to RAI-29.01 Re: License Amendment Request (LAR) to Adopt National Fire Protection Association Standard 805

4. Response to RAI-61 Re: License Amendment Request to Adopt National Fire Protection Association Standard 805

5. Response to RAI-62 Re: License Amendment Request to Adopt Nation Fire Protection Association Standard 805

6. Indiana Michigan Power Company National Fire Protection Association Standard 805 License Amendment Request Revisions Re: License Amendment Request to Adopt NFPA 805

7.

Donald C. Cook Nuclear Plant National Fire Protection Association Standard 805 Transition Report Section 5.4 Transition Implementation Schedule Re: License Amendment Request to Adopt NFPA 805

8. Revision 4 of Attachment M, "License Condition Changes," to the Transition Report Provided in Support of Response to RAI-29.01, 61 and 62 and Twelve Month Implementation Schedule

9. Revision 5 of Attachment S, "Plant Modifications and Items to be Completed During Implementation," to the Transition Report, Provided in Support of Response to RAI 29.01, 61 and 62 Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.

Enclosures 10 and 11 to this letter contain security-related information.

Upon removal of Enclosures 10 and 11, this letter is uncontrolled.

U. S. Nuclear Regulatory Commission Page 3 AEP-NRC-2013-17

10. Donald C. Cook Nuclear Plant National Fire Protection Association Standard 805 Transition Report Attachment C, NEI 04-02 Table B-3, Revision 1

11. Donald C. Cook Nuclear Plant National Fire Protection Association Standard 805 Transition Report Attachment G, Table G-1, Recovery Actions and Activities Occurring at the Primary Control Station(s), Revision 1 c:

C. A. Casto, NRC Region III J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosures MDEQ - RMD/RPS NRC Resident Inspector T. J. Wengert, NRC Washington, DC Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.

Enclosures 10 and 11 to this letter contain security-related information.

Upon removal of Enclosures 10 and 11, this letter is uncontrolled.

Enclosure I to AEP-NRC-2013-17 AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS k

DAY OF 2013 My Commissin EiNotars y--blic-My Commission Expires ****

Z\\0 DANIELLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 04-A4-2018 Acting in the County of L6oC--W\\

to AEP-NRC-2013-17 Identification of Documents Referenced in this Letter and Its Enclosures

References:

1. Letter from M. H. Carlson, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 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)," AEP-NRC-2011-1, dated July 1,2011, ADAMS Accession No. ML111188A145.

2. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Docket Nos. 50-315 and 50-316, Supplement to 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),"

AEP-NRC-2011-62, dated September 2, 2011, ADAMS Accession No. ML11256A030.

3. Letter from P. S. Tam, NRC, to L. J. Weber, I&M, "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),"

dated January 27, 2012, ADAMS Accession Nos.

ML113560709, ML12003A186, and ML12017A251.

4. E-Mail from P. S. Tam, NRC, to H. L. Etheridge, J. R. Waters, M. K. Scarpello, I&M, et al.,

"D.C. Cook - Draft RAI re. Transition to NFPA 805, Questions in Health Physics (TAC ME6629 and ME6630),"

dated March 22,

2012, ADAMS Accession No. ML12082A043.

5. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard

805, (TAC Nos.

ME6629 AND ME6630),"

AEP-NRC-2012-29, dated April 27, 2012, ADAMS Accession No. ML12132A390.

6. Letter from J. P Gebbie, I&M, to NRC Document Control Desk, "Response to Second Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805 (TAC Nos. ME6629 AND ME6630)," AEP-NRC-2012-47, dated June 29, 2012, ADAMS Accession No. ML12195A013.

7. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard

805, (TAC Nos.

ME6629 AND ME6630),"

AEP-NRC-2012-58, dated August 9, 2012, ADAMS Accession No. ML12242A246.

to AEP-NRC-2013-17 Page 2

8. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "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)," dated February 1, 2013, ADAMS Accession Nos.

ML13014A549.

9. E-Mail from T. Wengert, NRC, to H. L. Etheridge, J. M. Tanko, M. K. Scarpello, I&M, et al.,

"D.C. Cook - Draft RAI re. Transition to NFPA 805, UFSAR Description as a Result of Implementing NFPA 805 and (FAQ) 12-0062 Closure Memo, ADAMS Accession No. ML12082A043,"

dated March 20, 2013, (Identified as

RAI-62

in this letter AEP-NRC-2013-17).

10. I&M NFPA 805 LAR Revisions Action Request 2012-12753, dated October 12, 2012.

11. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "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)," dated October 11, 2012, ADAMS Accession Nos. ML12276A300 and ML12285A179.

12. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Second Round Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Associations Standard 805 (TAC Nos. ME6629 and ME6630),"

AEP-NRC-2012-92, dated October 15, 2012, ADAMS Accession No. ML122970218.

13. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Second-Round Request for Additional Information Item 54.b, and Submittal of Revised Tables Regarding the Application for Amendment to Transition to National Fire Protection Association Standard 805 (TAC Nos. ME6629 and ME6630)," AEP-NRC-2012-101, dated November 9, 2012, ADAMS Accession No. ML123261084.

14. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "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)," dated February 1, 2013, ADAMS Accession No. ML12345A327.

15. Letter from Q. S. Lies, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Third Round Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805 (TAC Nos. ME6629 and ME6630)," AEP-NRC-2013-1, dated January 14, 2013, ADAMS Accession No. ML13028A113.

to AEP-NRC-2013-17 Page 3

16. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Revised Response to a First Round Request for Additional Information Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805 (TAC Nos. ME6629 and ME6630),"

AEP-NRC-2013-09, dated February 1, 2013, ADAMS Accession No. ML13045A432.

to AEP-NRC-2013-17 Response to RAI-29.01 Re: License Amendment Request to Adopt National Fire Protection Association Standard 805 Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

This enclosure provides Indiana Michigan Power Company's (l&M's),

licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to Request for Additional Information (RAI) 29.01 from the U. S. Nuclear Regulatory Commission, transmitted by Reference 8, regarding the CNP transition to National Fire Protection Association Standard 805 (NFPA 805).

RAI 29.01 Probabilistic Risk Assessment The August 9, 2012, response to RAI-29 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12242A246) provides justification of a value of 0.1 to represent the failure to reach safe shutdown using alternate means. The justification largely consists of a qualitative argument that the seven feasibility assessment and considerations identified in NUREG-1921 are addressed by the Donald C. Cook Nuclear Plant (CNP) for alternate shutdown. A quantitative assessment of the failure of alternate shutdown is not presented. A comparison is made to human error probabilities (HEPs) associated with Station Blackout (SBO); however, it is not clear that the actions are of the same complexity. The cognition element of SBO actions may be more straightforward than actions following a fire event in the Main Control Room. Also, it appears that alternative shutdown consists of about 20 separate Recovery Actions involving approximately 50 different components. Given this level of complexity and potential number of actions, provide further justification for the 0. 1 HEP by providing the results of the human failure event (HFE) quantification process described in Section 5 of NUREG-1921, considering the following:

The results of the process in Section 5.2.7 of NUREG-1921 for assigning scoping HEPs to actions associated with switchover of control to an alternate shutdown location, specifically addressing the basis for the answers to each of the questions asked in the Figure 5.4 flowchart.

The results of the process in Section 5.2.8 of NUREG-1921 for assigning scoping HEPs to actions associated with the use of alternate shutdown, specifically addressing the basis for the answers to each of the questions asked in the Figure 5-5 flowchart.

Provide a sensitivity analysis that shows the impact on the PRA results (CDF, LERF, -CDF,

-LERF) of using the NUREG-1921 Section 5.2 HEP(s) for control room abandonment scenarios. Alternatively, provide the results of a detailed human reliability analysis (HRA) quantification, per Section 5.3 of NUREG-1921, if the screening conditional core damage probability (CCDP) of 0.1 and the NUREG-1921 Section 5.2 scoping HEPs are determined to not be representative.

to AEP-NRC-2013-17 Page 2 Response to RAI-29.01 In the Transition Report, the CNP Fire Probabilistic Risk Assessment (PRA) used a single, overall HEP of 0.1 as representative of the set of operator actions needed to safely shutdown the plant following a fire in the Main Control Room (MCR) or a fire in the MCR cable vault. The single HEP was applied to all MCR and MCR cable vault fire scenarios that led to evacuation, whether caused by loss of habitability or loss of system control. RAI-29 and RAI-29.01 asked for further justification for applying a single representative HEP to the range of scenarios that would be encountered, each of which potentially involves many operator actions. This response evaluates detailed HEPs for each required human error event and applies them explicitly to each scenario.

Incorporating detailed HEPs for MCR evacuation scenarios into the CNP Fire PRA model resulted in the delta-Core Damage Frequency (CDF) limit of Regulatory Guide (RG) 1.205 being exceeded slightly. CNP is in the process of replacing the existing Reactor Coolant Pump (RCP) seals with a new, low-leakage version of the seal which will reduce the mitigation requirements of an RCP seal Loss of Coolant Accident. The analysis using detailed HEPs for MCR evacuation scenarios was then re-quantified including credit for the low-leakage RCP seals in the Fire PRA model.

A discussion of the HEP modeling changes and the results of incorporating both the detailed HEPs and low-leakage RCP seals are provided in this response.

MCR HEP DEVELOPMENT:

All MCR evacuation scenarios for Units 1 and 2 were revisited in this effort. Scenario-specific HEP values were developed for each HFE. The HEPs for ex-control room actions were re-evaluated using the most recent feasibility analysis performed by CNP as an input. All of the operator actions were evaluated using detailed HRA methods, per Section 5.3 of NUREG-1921.

Fault trees representing system and equipment response during the MCR evacuation scenarios were solved to obtain overall evacuation CCDP values. These fault trees allowed quantification of Large Early Release Frequency (LERF) in a manner similar to the non-evacuation CDF scenarios in the Fire PRA.

Calculation PRA-FIRE-17663-012-RAI-29, CNP Fire PRA Human Reliability Analysis Update to Main Control Room Abandonment HRA, documents that sensitivity case as requested in RAI-29.01.

In addition, evacuation probabilities for several classes of panel fires were updated using results from CFAST software analyses that had been revised in accordance with earlier RAI responses for RAIs 46 - 52 that are found in Reference 12.

The recovery actions for each scenario can be grouped into two categories:

1. Actions Needed for All Scenarios. This category of actions consists of those required to restore auxiliary feedwater (AFW) and restore borated injection using Chemical and Volume Control System (CVCS), each with associated support systems. These are the minimum set of systems necessary to provide safe shutdown. These actions include aligning AFW from the opposite unit, aligning CVCS from the opposite unit, and to AEP-NRC-2013-17 Page 3 establishing instrumentation power from the other unit. Failure to provide any of these actions is modeled as leading to core uncovery. These actions are required for all MCR evacuation scenarios and are modeled as a single basic event in the Fire PRA. The detailed analysis of these actions includes models of cognitive errors and execution errors following the NUREG-1921 guidelines.

2. Additional Actions Needed to Mitigate Spurious Actuations. This category consists of actions required to mitigate spurious equipment actuation and restore the Reactor Coolant System (RCS) and Steam Generator (SG) boundaries to a state where the AFW and CVCS systems can provide for safe shutdown. This category of actions includes RCP trip, isolation of RCS boundary valves (i.e., pressurizer power operated relief valves (PORVs), RCS head vent, pressurizer vent, and RCS letdown), isolation of the SGs (i.e., closure of open main steam safety valves, closure of open SG PORVs, and closure of open SG blowdown valves), and termination of spurious safety injection.

These actions are required only when fire damage causes a spurious event which must be terminated.

The HFEs for these events are modeled in cutsets along with the component failure modeling the spurious event. The detailed analysis of these actions includes models of cognitive errors and execution errors following the NUREG-1921 guidelines.

RISK RESULTS:

Table 1 shows the risk results submitted in the NFPA-805 Transition Report (References 1 and

2) side by side with the results for the CNP Fire PRA model that has been revised to include detailed HEPs for MCR evacuation scenarios and low-leakage RCP shutdown seals. Table 2 shows the same comparison for Unit 2.

TABLE 1: RESULTS FOR UNIT I MCR EVACUATION SCENARIOS USING REVISED HEPs AND LOW-LEAKAGE RCP SEALS Transition Report Risk Risk Values with Revised MCR MCR Fire Scenarios Values (lyr)

HEPs & RCP Seals (/yr)

Requiring Evacuation CDF LERF CDF LERF MCR Back Panels 6.16E-7 8.59E-8 1.55E-7 1.83E-8 Main Control Boards 3.90E-9 5.42E-10 2.24E-9 2.56E-10 MCR Cable Vault 1.36E-6 1.78E-7 1.12E-6 1.29E-7 Total MCR Evacuation 1.98E-6 2.64E-7 1.28E-6 1.48E-7 Scenario Risk Change in Delta Risk from

-7.OOE-7

-1.16E-7 Transition Report Unit 1 Total Delta Risk with Only MCR Evacuation Scenario Changes 9.01 E-6 6.85E-7 8.31E-6 5.69E-7 (remainder of areas from Transition Report) to AEP-NRC-2013-17 Page 4 TABLE 2: RESULTS FOR UNIT 2 MCR EVACUATION SCENARIOS USING REVISED HEPs AND LOW-LEAKAGE RCP SEALS Transition Report Risk Risk Values with Revised MCR MCR Fire Scenarios Values (/yr)

HEPs & RCP Seals (/yr)

Requiring Evacuation CDF LERF CDF LERF MCR Back Panels 6.16E-7 8.59E-8 1.62E-7 1.82E-8 Main Control Boards 3.90E-9 5.42E-10 2.24E-9 2.52E-10 MCR Cable Vault 1.OOE-6 1.36E-7 1.19E-6 1.52E-7 Total MCR Evacuation 1.62-6 2.22E-7 1.35E-6 1.70E-7 Scenario Risk Change in Delta Risk from

-2.70E-7

-5.20E-8 Transition Report Unit 2 Total Delta Risk with Only MCR Evacuation Scenario Changes 8.46E-6 5.97E-7 8.19E-6 5.45E-7 (remainder of areas from Transition Report)

DISCUSSION OF RESULTS:

The net result of incorporating detailed HEP values for MCR evacuation fire scenarios and crediting the RCP shutdown seals shows a net decrease in core damage frequency of 7.O0E-7/2.70E-7 per year in delta CDF for Unit 1/Unit 2, respectively, and 1.16E-7/5.20E-8 per year in delta LERF for Unit 1/Unit 2, respectively.

These values for both units meet the RG 1.205 acceptance criteria. I&M has revised Attachment S, Table S-2, to add an implementation item (Item S-2.3) for the planned modification to upgrade the existing RCP seals to the new, low-leakage design. Additionally, the RCP seal modification has been added to the license conditions in Attachment M to the Transition Report.

to AEP-NRC-2013-17 Response to RAI-61 Re: License Amendment Request to Adopt National Fire Protection Association Standard 805 Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

This enclosure provides Indiana Michigan Power Company's (l&M's),

licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, revised response to Request for Additional Information (RAI) 61 from the U. S. Nuclear Regulatory Commission, transmitted by Reference 8, regarding the CNP transition to National Fire Protection Association Standard 805.

RAI 61 Probabilistic Risk Assessment The responses to the following RAIs provided sensitivity analyses to show the impact on fire risk of the indicated PRA modeling, and this included PRA upgrades to meet Capability Category II (CC-Il) requirements:

RAI 20 regarding compliance with LERF-related supporting requirements in the PRA standard RAI 29. 01 regarding the use of a CCDP of 0. 1 RAI 38 regarding credit for control power transformers RAI 31 regarding propagation of parametric uncertainty RAIs 34.e and 34.01 regarding transient fire frequency apportionment RAI 40.c, Sensitivity Case 8, regarding treatment of transient fires near pinch points in inaccessible locations The Nuclear Regulatory Commission staff's position is that each of these sensitivity analyses:

1) has a potentially significant impact on the delta risk for the plant based on the results of the individual sensitivity analysis, and 2) if they are in the post-transition plant base case, must be evaluated collectively to assure a comprehensive treatment of the CNP transition fire risk.

Provide the results of a composite sensitivity analysis that shows the integrated impact on the fire risk (Core Damage Frequency (CDF), Large Early Release Frequency (LERF), ACDF, ALERF) of all of the above sensitivity analyses in the post-transition plant base case. In this composite analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. In the response, provide justification of how the Regulatory Guide (RG) 1.174 risk acceptance guidelines are satisfied for the composite sensitivity analysis of all the issues identified in this RAI and, if applicable, a description of any new modifications or operator actions being credited to reduce delta risk and the associated impacts to the fire protection program. If the licensee elects to exclude any of the above sensitivity analyses in the composite analysis, provide justification for exclusion. In addition, the licensee may provide an additional composite sensitivity analysis that includes only the sensitivity analyses that it believes appropriate. Again, provide justification of how the RG 1.174 risk acceptance guidelines are satisfied.

to AEP-NRC-2013-17 Page 2 Response to RAI-61 The CNP Fire Probabilistic Risk Assessment (PRA) model has been revised to include resolutions of the following RAIs:

RAI-20: Revise LERF split fractions in response to an internal events PRA model gap assessment.

RAI-29 & 29.01: Use scenario specific human error probabilities (HEPs) for main control room (MCR) evacuation scenarios.

RAI-31: Perform correlated uncertainty analysis on the combined core damage equations to derive a calculated mean value.

RAI-34e: Redistribute the transient fire frequency from NUREG/CR-6850 in accordance with the weighting factors and distribution scheme developed in FAQ-12-0064.

RAI-38: Eliminate credit for existence of a Control Power Transformer (CPT) to reduce the probability of a spurious component operation. For the current model development, the value of 0.5 is used.

RAI-40c, Sensitivity B: For fire areas AA48 and AA52 (switchgear rooms and cable vaults) transient fires must be assumed to occur in all locations. An occupancy split fraction of 0.33 times the floor area weighted transient fire frequency was used for inaccessible locations.

When these RAI resolutions are incorporated into the PRA model as listed above, the delta CDF and delta LERF for NFPA 805 transition do not meet the RG 1.205 criteria. As a result, I&M has undertaken two changes to improve risk. I&M has committed to install new low-leakage Reactor Coolant Pump (RCP) Seals and to credit a recovery action for the turbine driven auxiliary feed water (AFW) pump on loss of DC power. This response provides the risk metrics for CDF and LERF for the revised aggregate sensitivity model, with the two risk improvements listed above.

In addition, delta risk was separated from deterministic risk for MCR sequences.

Calculation PRA-FIRE-17663-015-RAI-61: CNP Fire PRA Uncertainty and Sensitivity Analyses documents the aggregate sensitivity case as requested in RAI-61.

to AEP-NRC-2013-17 Page 3 The specific treatment of each issue is summarized in Table 1 below.

Table 1: Disposition of Each RAI Issue for Model Update RAI MODEL MODIFICATIONS

RAI-20

1. LERF5 was replaced with LERF5A1 or LERF5B1. LERF5B1 was used for designated station blackout scenarios. All other LERF5 cases were changed to LERF5A1.

2. Values for the LERF split fractions for LERF2, LERF4, LERF5B1, LERF5A1, and LERF6 were changed to be consistent with the RAI-20 response.

RAI-29 and RAI 29.01

1. Separate evacuation probabilities for MCR panel fires were (MCR Re-analysis) developed based on characteristics of the panels and location.

The cabinet types are split into open/closed and unqualified/qualified cable. Each case has its own evacuation probability.

2. Use of the 0.1 HEP for MCR evacuation was discontinued. Event specific human reliability assessment was provided for recovery actions during evacuation scenarios. The human failure events with their specific HEPs were explicitly modeled in the fault trees.

The scenario Conditional Core Damage Probabilities (CCDPs) were quantified using scenario specific requirements for recovery actions.

3. Recovery actions for a) closure of a spurious open Steam Generator (SG) Pressure Operated Relief Valve (PORV) and b) closure of a main steam stop valve was previously set to a screening value of 0.1. These were evaluated in detail and now have an HEP of 9.4E-3.

4. Alignment of instrumentation to monitor Reactor Coolant System (RCS) and SG levels and temperatures is now quantified explicitly. The HEP is assigned 4.4E-3 for SG water level instrumentation and 4.4E-3 for pressurizer water level instrumentation.

5. CCDP values for evacuation scenarios were re-run to allow component failures in the specific scenarios to trigger related recovery actions.

6. MCR scenarios were split into delta risk and deterministic risk.

Previously all CDF from MCR scenarios was considered delta risk. Now, only those scenarios with a variance from deterministic requirements are considered delta risk.

7. CDF from back panel fires which are suppressed with no safe shutdown component damage are quantified and assigned to deterministic risk.

to AEP-NRC-2013-17 Page 4 Table 1: Disposition of Each RAI Issue for Model Update RAI MODEL MODIFICATIONS

RAI-30

A recovery action was included to operate the turbine driven AFW pump after battery depletion in the cases of load shed of the battery charger after a Safety Injection or Loss of Offsite Power signal on the 4kV safety busses. This action represents re-alignment of the battery charger for those scenarios in which a 4kV safety bus is available.

For station blackout scenarios, this recovery action represents manual local control of AFW pump discharge valves and local reading of SG water level.

RAI-31

Parametric uncertainty was performed on the aggregated CDF and LERF equations. This was run using a Monte Carlo process inside the WinNUPRA PRA software. All probabilities derived from the same data and exhibiting the same probability value were correlated.

Correlation of the spurious probabilities causes a significant increase in the calculated mean. All basic spurious valve events with a 0.5 probability were correlated. The distribution was chosen as a gamma distribution with a mean of 0.5 and a variance of.02. These parameters result in a range factor of about 1.65 and a 95%

percentile of.76; a 99% percentile of.90 and a 99.9% of 1.07.

RAI-34e FAQ 12-0064 was used to redistribute transient fire frequency for the entire CNP (i.e., both units combined).

RAI-38

The probability of spurious operation for all components which previously credited a CPT to reduce spurious event probability was changed to 0.5. This probability was used for spurious open and spurious close.

RAI-40c, Sensitivity B RAI-40c questioned the possibility that transient fires could be located in any location in Fire Area AA48 and AA52, regardless of the perceived inaccessibility of the location. Previously, for the switchgear rooms and auxiliary cable vaults, certain floor areas had been postulated to be transient free. The room transient fire frequency was distributed over the remainder of the floor location. For sensitivity study B, the "inaccessible" locations were assigned one third (0.33) of their fair transient frequency based on floor area. Rather than performing fire modeling for these new transient fires, the CCDP/Conditional Largest Early Release Probability (CLERP) for these transient fires was assigned the worst case value of any other fire in the room (which happens to be the whole room burnout CCDP/CLERP).

The CNP Fire PRA has been modified to incorporate all changes discussed above. This revised model will become the base-case Fire PRA model for all fire risk related activities at the completion of the 12 month Transition Period.

All scenarios for both units have been to AEP-NRC-2013-17 Page 5 requantified using the new base case model. The CDF, LERF, delta CDF, and delta LERF values are all significantly reduced over the corresponding values reported in the Transition Report (Reference 1).

The CDF/LERF reduction is primarily due to the credit for the low-leakage RCP seals to prevent significant seal leakage upon loss of RCP seal cooling. This functional development significantly reduced the risk importance of recovery actions to trip of RCPs. It also diminished the risk importance of recovery actions to restore RCS inventory makeup.

The second most important risk reduction change was to implement a recovery action for the turbine driven AFW pump after battery depletion.

The point estimate results are shown in Table 2 for both units. The uncertainty ranges for the risk metrics are shown in Table 3 for Unit 1 and Table 4 for Unit 2. Note that the Internal Events PRA CDF and LERF values are comparable to the corresponding Fire PRA values shown, while the current estimates for Seismic Risk are less than the Fire PRA values shown.

Table 2: Risk Metrics of the CNP Fire PRA with New, Low-Leakage RCP Seals PARAMETER UNIT I UNIT 2 RG 1.174 Criteria TOTAL FIRE 1.13E-5 1.43E-5 Total CDF for all PRA Initiating CDF Events <1 E-4/yr (per year)

TOTAL FIRE 1.89E-6 2.11E-6 Total LERF for all PRA Initiating LERF Events <1 E-5/yr (per year)

DELTA CDF 2.19E-6 4.04E-6 Risk from Fire <1E-5/yr (per year)

DELTA LERF 3.12E-7 6.1OE-7 Risk from Fire <1E-6/yr (per year) to AEP-NRC-2013-17 Page 6 Table 3: Correlated Uncertainty Calculation for CNP Unit I Risk Metrics METRIC 5TH 5 0 TH Point MEAN 9 5 TH Percentile Percentile Estimate Percentile Unit 1 Fire CDF 4.96E-6 9.71 E-6 1.13E-5 1.12E-5 2.22E-5 Unit 1 Fire LERF 8.22E-7 1.50E-6 1.89E-6 1.78E-6 3.60E-6 Unit 1 Delta CDF 1.OOE-6 1.96E-6 2.19E-6 2.36E-6 4.94E-6 Unit 1 Delta LERF 1.68E-7 2.90E-7 3.12E-7 3.20E-7 5.83E-7 Table 4: Correlated Uncertainty Calculation for CNP Unit 2 Risk Metrics METRIC 5TH 50TH Point MEAN 95TH Percentile Percentile Estimate Percentile Unit 2 Fire CDF 7.99E-6 1.18E-5 1.43E-5 1.30E-5 2.1OE-5 Unit 2 Fire LERF 1.32E-6 1.89E-6 2.11E-6 2.05E-6 3.21E-6 Unit 2 Delta CDF 2.29E-6 4.08E-6 4.04E-6 4.43E-6 7.74E-6 Unit 2 Delta LERF 2.99E-7 5.66E-7 6.1OE-7 6.26E-7 1.15E-6 DISCUSSION OF RESULTS:

The quantification of the new base case Fire PRA model, developed in response to RAI-61, "Aggregate Sensitivity Study," combined with crediting the new low-leakage RCP shutdown seals and turbine-drive AFW pump recovery shows the transition risk metrics for Unit 1 and for Unit 2, including the total delta-risk for each CNP unit to be below the RG 1.205 delta-risk thresholds.

This is true for both the point estimate results as well as the correlated mean values. I&M has revised Attachment S, Table S-2, to add an implementation item (S-2.3) for the planned modification to upgrade the existing RCP seals to the new low-leakage design.

Additionally, the RCP seal modification has been added to the license conditions in Attachment M to the Transition Report.

The response to RAI-61 develops the Fire PRA risk metrics for a collective set of RAIs that were previously evaluated individually. Additionally, the RAI-61 risk metrics also include the risk reduction due to the plant modification to replace the RCP seals. While these changes alter the risk profile somewhat from that presented in Attachment W of the Transition Report, meaning the risk contributions of individual areas may change, the aggregate set of FPRA changes combine to produce an overall reduction of each of the RG 1.174 risk metrics (CDF, LERF, delta CDF, and delta LERF values) used for decision-making.

to AEP-NRC-2013-17 Response to RAI-62 Re: License Amendment Request to Adopt National Fire Protection Association Standard 805 Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

This enclosure provides Indiana Michigan Power Company's (l&M's),

licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to Request for Additional Information (RAI) 62 from the U. S. Nuclear Regulatory Commission, transmitted by Reference 9 regarding the CNP transition to National Fire Protection Association Standard 805.

RAI 62 UFSAR Description as a Result of Implementing NFPA 805 and FAQ 12-0062 Closure Memo Section 4.6 of Nuclear Energy Institute (NEI) 04-02 indicates that the LAR should contain a "discussion of the changes to Updated Final Safety Analysis Report (UFSAR) necessitated by the license amendment and a statement that the changes will be made in accordance with 10 CFR 50.71(e)." LAR Attachment S, Implementation Item S-3.13, indicates that UFSAR Section 9.8.1 will be revised as needed for implementation of NFPA 805. However, there is no description of the changes that need to be made to the current UFSAR.

Describe the changes that will to be made to the current UFSAR as a result of implementing NFPA 805. Alternatively, indicate whether the UFSAR will be updated following the guidance provided in the Frequently Asked Question (FAQ) 12-0062 closure memo (ADAMS Accession No. ML121980557).

Response to RAI-62 I&M has revised Attachment S, Table S-3, Implementation Item S-3.13 to state, "update the UFSAR following the guidance provided in Frequently Asked Question (FAQ) 12-0062 (ADAMS Accession No. ML121430035)."

to AEP-NRC-2013-17 Indiana Michigan Power Company National Fire Protection Association Standard 805 License Amendment Request Revisions Re: License Amendment Request to Adopt NFPA 805 Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

During preparation of Indiana Michigan Power Company (I&M) letter dated October 15, 2013 (Reference 12), discrepancies were noted in the risk related values provided in Table W2, "Fire Initiating Event Scenarios Contributing More than 1% of the Calculated Fire Risk for Unit 2," in the original amendment request (Reference 1). To resolve the discrepancies, a revised Table W2 was provided by Reference 13.

The discrepancies were entered into I&M's Corrective Action Program as Action Request 2012-12753 and an extent of condition evaluation was completed.

Engineering evaluation of the discrepancy determined the inconsistencies were inconsequential with respect to impact on the Donald C. Cook Nuclear Plant (CNP) National Fire Protection Association Standard 805 (NFPA 805) Transition Report. However, the extent of condition evaluation identified additional instances within the original CNP NFPA 805 Transition Report (Reference 1) and supporting documentation that require corrective action. The Action Request documenting the inconsistencies with their respective cause and corrective action has been posted to the Sharepoint that has been utilized for documents and information exchange between I&M and the Nuclear Regulatory Commission staff for the License Amendment Request to Adopt NFPA 805.

Attachment C and Attachment G of the CNP NFPA 805 Transition Report required revision.

Accordingly, Enclosure 10 of this letter identifies and provides the revised pages of the CNP NFPA 805 Transition Report, Attachment C, NEI 04-02 Table B-3, Revision 1. Enclosure 11 of this letter identifies and provides revised pages of the CNP NFPA 805 Transition Report, Attachment G, Table G-1, Recovery Actions and Activities Occurring at the Primary Control Station(s), Revision 1.

Aggregate review of the inconsistencies concluded the discrepancies are inconsequential with respect to adverse impact on the original CNP NFPA 805 License Amendment Request.

to AEP-NRC-2013-17 Donald C. Cook Nuclear Plant National Fire Protection Association Standard 805 Transition Report Section 5.4 Transition Implementation Schedule Re: License Amendment Request to Adopt NFPA 805 Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

Indiana Michigan Power Company (I&M) requests an additional six months from the original National Fire Protection Association Standard 805 (NFPA 805) Transition Report Implementation Schedule submitted by References 1 and 2, (ref. CNP NFPA 805 Transition Report, Revision 0, Section 5.4, Page 115). Revision 0 of the Donald C. Cook Nuclear Plant (CNP) NFPA 805 Transition Report, Section 5.4, and Attachment S, Table S-3, also states implementation of the new NFPA 805 fire protection program "within six (6) months".

This enclosure provides the CNP NFPA 805 Transition Report Section 5.4, Page 115. Enclosure 9 to this submittal includes the revised Attachment S, Table S-3. Both documents reflect implementation of NFPA 805 "within twelve (12) months" after Nuclear Regulatory Commission (NRC) approval and issuance of the NRC Safety Evaluation (SE).

Justification is based on site resources required to prepare for and support 2013 CNP activities including Unit 1 and Unit 2 spring and fall Refueling Outages, April 2013 NRC License Renewal Inspection Phase I, May-June 2013 Fire Protection Triennial Inspection, and August 2013 INPO Evaluation.

5.4 Transition Implementation Schedule I&M's schedule for transition of CNP to the new fire protection licensing basis is as follows:

Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel. This will occur within twelve (12) months after issuance of the NFPA 805 SE.

Attachment S provides a listing of plant modifications associated with the transition to NFPA 805. I&M will initiate the implementation of plant modifications following submittal of the License Amendment Request and anticipates completion of installation in the plant within twelve (12) months after issuance of the NFPA 805 SE.

Revision 1 of the CNP NFPA 805 Transition Report, Section 5.4, Transition Implementation Schedule with the revised wording is attached.

Indiana Michiaan Power CNP NFPA 805 Transition ReDort 5.4 Transition Implementation Schedule I&M's schedule for transition of CNP to the new fire protection licensing basis is as follows:

Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel. This will occur within twelve (12) months after issuance of the NFPA 805 SE.

" Attachment S provides a listing of plant modifications associated with the transition to NFPA 805. I&M will initiate the implementation of plant modifications following submittal of the LAR and anticipates completion of installation in the plant within twelve (12) months after issuance of the NFPA 805 SE.

Revision 1 Page 115 Revision 1 Page 115 to AEP-NRC-2013-17 Revision 4 of Attachment M, "License Condition Changes," to the Transition Report Provided in Support of Response to RAI-29.01, 61 and 62 and Twelve Month Implementation Schedule Changes are indicated by revision bars in the right margin.

Indiana Michiaan Power CNP NFPA 805 Transition ReDort - Attachment M Inin ihia Poe N

FA85 rniinRort-Atcmn M. License Condition Changes 5 Pages Revision 4 Page M-1 Revision 4 Page M-1

CNP NFPA 805 Transition Rer)ort - Attachment M IJIndiana*

Michina Polg wer I&M proposes to replace the current CNP fire protection license conditions 2.C.(4) for Unit 1 and 2.C.(3)(o) for Unit 2 with the standard license condition in Regulatory Position C.3.1 of Regulatory Guide 1.205, Revision 1, as shown below.

In support of this change, I&M has developed a Fire PRA which has been reviewed and been found acceptable by a Fire PRA WOG peer review conducted during October 12-16, 2009. Outstanding high level findings from the peer review are included in Attachment V of this TR. Any future changes to the Fire PRA will be subject to peer review in accordance with the guidance provided in NEI 07-12 and applicable ASME/ANS PRA standards.

Indiana Michigan Power shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated (and supplements dated

) and as approved in the Safety Evaluation Report dated (and supplements dated

). Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met.

The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(a)

Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b)

Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1X10 7/year (yr) for CDF and less than 1X10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval (1)

Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an Revision 4 Page M-2

CNP NFPA 805 Transition Report - Attachment M UUUIndan ichna PoIV erIIU I

engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard."

Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

"Fire Alarm and Detection Systems" (Section 3.8);

"Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9);

"Gaseous Fire Suppression Systems" (Section 3.10); and, "Passive Fire Protection Features" (Section 3.11).

(2)

Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact.

The licensee may use its screening process as approved in the NRC Safety Evaluation Report dated to determine that certain fire protection program changes meet the minimal criterion.

The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions (1)

Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2)

The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by (a)

Modify the CO 2 system from manual to automatic actuation in the following fire areas:

AA40 AA43 Revision 4 Page M-3

Indiana Michiaan Power CNP NFPA 805 Transition Report - Attachment M (b)

Modify motor operated valves to preclude IN 92-18 conditions in which control circuit short circuits can occur between control wiring and power sources leading to spurious operation of the valve, and in which the same postulated short circuit may bypass the torque/limit switches which, combined with the absence or bypass of thermal overload contacts, can result in continuous energization of the valve motor and potential mechanical damage to the valve such that manual operation via handwheel would be inhibited. The affected valves are identified in Attachment S.

(c)

Modify the Unit 1 and 2 Reactor Coolant Pump seals with the new low-leakage safe shutdown seals (SDS).

(3)

The Licensee shall implement the items listed in Enclosure (2) Table S-3 of the license amendment request within 365 days of issuance of the Safety Evaluation Report.

(4)

The Licensee shall complete an FPRA Focused Scope Peer Review and resolve findings associated with the revised FPRA LERF values, prior to self-approval of changes that result in more than a minimal increase in risk.

(5)

The Licensee shall complete an FPRA Focused Scope Peer Review and resolve findings associated with FPRA modeling of the installation of the new RCP Low-Leakage Safe Shutdown Seals, prior to self-approval of changes that result in more than a minimal increase in risk.

(6)

The Licensee shall complete a Focused Scope Peer Review and resolve findings of the PRA upgrade related to Reduced Mission Times for cutsets containing a Test and Maintenance event combined with a running failure, prior to self-approval of changes that result in more than a minimal increase in risk.

The license conditions to be replaced are restated below.

License condition 2.C(4) for Unit 1:

Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31, 1979, January 30, 1981, February 7, 1983, November 22, 1983, December 23, 1983, March 16, 1984, August 27, 1985, June 30, 1986, January 28, 1987, May 26, 1987, June 16, 1988, June 17, 1988, June 7, 1989, February 1, 1990, February 9, 1990, March 26, 1990, April 26, 1990, March 31, 1993, April 8, 1993, December 14, 1994, January 24, 1995, April 19, 1995, June 8, 1995, and March 11, 1996, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Revision 4 Page M-4

CNP NFPA 805 Transition Report - Attachment M Indiana Michinan Po er License condition 2.C(3)(o) for Unit 2:

Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31, 1979, January 30, 1981, February 7, 1983, November 22, 1983, December 23, 1983, March 16, 1984, August 27, 1985, June 30, 1986, January 28, 1987, May 26, 1987, June 16, 1988, June 17, 1988, June 7, 1989, February 1, 1990, February 9, 1990, March 26, 1990, April 26, 1990, March 31, 1993, April 8, 1993, December 14, 1994, January 24, 1995, April 19, 1995, June 8, 1995, and March 11, 1996, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if these changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

It is I&M's understanding that, implicit in the replacement of these license conditions, all prior fire protection program SERs and commitments will be superseded in their entirety by the revised license condition.

No other license conditions need to be replaced or revised.

I&M implemented the following process for determining that these are the only license conditions required to be either revised or superseded to implement the new fire protection program which meets the requirements in 10 CFR 50.48(a) and 50.48(c):

A review was conducted of the I&M Unit 1 Renewed License Number DPR-58, through Amendment No. 313 and Unit 2 Renewed License Number DPR-74, through Amendment No. 297, by the I&M licensing and NFPA 805 Transition Team. Outstanding LARs that have been submitted to the NRC but not yet approved were also reviewed for potential impact on the license conditions.

Revision 4 Page M-5 Revision 4 Page M-5 to AEP-NRC-2013-17 Revision 5 of Attachment S, "Plant Modifications and Items to be Completed During Implementation," to the Transition Report Provided in Support of Response to RAI 29.01, 61 and 62 and Twelve Month Implementation Schedule.

Changes are indicated by revision bars in the right margin.

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Inrlinnn Minhirinn Pnwor I-AIID AIX=JDA Pf)g Trnnoifinn m nrf-Aftof-hmont Q Tables S-1 and S-2, Plant Modifications, include a description of the modifications along with the following information:

A problem statement, Risk ranking of the modification, An indication if the modification is currently included in the Fire PRA, Compensatory Measure in place, and A risk-informed characterization of the modification and compensatory measure.

The following legend applies to the risk ranking indicated in Tables S-1 and S-2:

High = Modification would have an appreciable impact on reducing overall fire CDF.

Medium = Modification would have a measurable impact on reducing overall fire CDF.

Low = Modification would have either an insignificant or no impact on reducing overall fire CDF.

Revision 5 Page S-2 Revision 5 Page S-2

Incliana Mirhiaan Pnwpr Indina Mchien PwerCNP NFPA 805 Transition Rtnnnd - Attachment S Attachment S - Table S-1 Plant Modifications Completed In Comp Risk Informed Item Rank Unit Problem Statement Proposed Modification FPRA Measure Characterization S-1.1 High 1, 2 Cable in conduit associated Provide 1-hour ERFBS with Y

N Fire PRA credits this with the credited train of dc automatic suppression &

modification for electrical electrical power for Fire Area detection for cable of power redundancy AA39A and AA45A is concern in Fire Area AA39A Compensatorv measure:

unprotected and routed and AA45A None; modification through the area installed.

S-1.2 Low 1, 2 Actions identified in safe Revised safe shutdown N

N Fire PRA does not credit shutdown procedure for procedures to reflect correct these actions transferring 600 volt Bus to actions.

alternate source in Fire Areas Compensatory measure:

AA14, AA23, AA39B and None; Procedures AA45B were not accurately updated.

identified.

Revision 5 Page S-3 Revision 5 Page S-3

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-Afhen Attachment S - Table S-2 Plant Modifications Committed Proposed In Comp Risk Informed Item Rank' Unit Problem Statement Modification FPRA Measure Characterization S-2.1 Medium 1, 2 Short circuits can occur in Modify the following ten Y

N Circuit failures which result in the the control circuit for (10) valves to resolve inability of operators to perform a motor operated valves 92-18 issue:

recovery action are significant between control wiring 1(2)-FMO-212 contributors to both fire risk as well and power sources as internal events risk. The leading to spurious 1(2)-FMO-222 proposed modification will improve operation of the valve, the ability of plant operators to The same postulated 1(2)FMO232 manually align certain motor short circuit may bypass 1(2)-FMO-242 operated valves if required in order the torque/limit switches to maintain the plant in a safe and which, combined with the 1(2)-ICM-250 stable condition.

absence or bypass of thermal overload Compensatory measure for NFPA contacts results in 805: None; This modification will be continuous energization completed prior to implementation of the valve motor and of NFPA 805 FP Program..

potential mechanical Compensatory measure for damage to the valve such 10 CFR 50 Appendix R: None; IN that manual operation via 92-18 concerns were addressed by handwheel would be the current licensing basis, which inhibited. This potential credits the "double break" circuit condition was described design, as described in NRC SE in Information Notice (IN) dated November 22, 1983.

92-18 Revision 5 Page S-4 Revision 5 Page S-4

Indiana Mit'ohirn Prower C lP NIZPA 805 Trmvzitinn Pnnrt - Attac'hmeVnt S Attachment S - Table S-2 Plant Modifications Committed Proposed In Comp Risk Informed Item Rank Unit Problem Statement Modification FPRA Measure Characterization S-2.2 High 1, 2 Electrical cabinet fire Modify the 002 system Y

N This proposed modification will limit scenarios in fire areas from manual to the extent of damage predicted to AA40 and AA43 automatic actuation in occur for fire scenarios within the contribute significantly to the following fire areas:

subject fire areas.

fire CDF and LERF and warrant additional a) AA40 Compensatory measure for NFPA mitigation b) AA43 805: None; This modification will be completed prior to implementation of NFPA 805 FP Program.

Compensatory measure for 10 CFR 50 Appendix R: None; fire areas AA40 and AA43 are deterministically compliant with 10 CFR 50 Appendix R.

S-2.3 High 1,2 Additional risk reduction Modify the Unit 1 and 2 Y

N This proposed modification will modifications are Reactor Coolant Pump provide an FPRA risk reduction necessary to comply with seals with the low-within the acceptance guidelines of Regulatory Guide 1.174 leakage seal design Regulatory Guide 1.174 acceptance guidelines Compensatory measure for NFPA 805: None; This modification will be completed prior to implementation of NFPA 805 FP Program.

Compensatory measure for 10 CFR 50 Appendix R: None; fire areas AA56 and AA58 are deterministically compliant with 10 CFR 50 Appendix R.

Revision 5 Page S-5 Revision 5 Page S-5

Indiana Mich4qan Power CNP NFPA 805 Transition Report - Attachment S Table S-3, identifies those implementation items (procedure changes, process updates, and training to affected plant personnel) that will be completed by I&M prior to the implementation of new NFPA 805 FP program. These items will be completed within twelve (12) months after NRC issuance of the NFPA 805 SE.

Revision 5 Page S-6

Indiana Michician Power CNP NFPA 805 Transition ReDort - Attachment S Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section I Source S-3.1 1, 2 Initial General Employee Training (GET) will be verified and / or 4.1.2 and Attachment A updated to include the minimum fire protection program elements as discussed in Section K to NEI-04-02 (FAQ 06-0028).

S-3.2 1,2 The monitoring program required by NFPA 805 Section 2.6 will be 4.1.2, 4.6, and Attachment A developed in accordance with NFPA 805 FAQ 10-0059, and will include a process that reviews the FPP performance and trends in performance and implemented after the LAR approval as part of the FPP transition to NFPA 805.

S-3.3 1, 2 Transient Combustible Free Zones will be established in high risk 4.5 and Attachments A and W Fire Areas AA40, AA43, AA48, AA50, AA51, and AA52.

S-3.4 1, 2 Hot Work Restriction Zones will be established in high risk Fire 4.5 and Attachments A and W Areas AA40, AA43, AA48, AA50, AA51, and AA52.

S-3.5 1, 2 Post-fire operating procedures will be updated to reflect new 4.2.1.3 and Attachment G NSCA strategies and training performed as necessary.

S-3.6 1, 2 Technical and administrative procedures and documents that 4.3.2 and Attachment D relate to non-power modes of plant operating states will be revised as needed for implementation of NFPA 805.

Revision 5 Page S-7 Revision 5 Page S-7

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w~,i UV~'h~i MIZ Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section / Source S-3.7 1, 2 Pre-fire plans and fire brigade training materials will be revised to 4.4.2 and Attachments A and E reflect changes required to meet the NFPA 805 radioactive release performance criteria.

S-3.8 1, 2 A confirmatory demonstration (field verification walk-through) of 4.2.1.3 and Attachment G the feasibility for the credited NFPA 805 recovery actions will be performed. This will include field verification of:

(1) Transit times (i.e., travel times to/from recovery action manipulated plant equipment).

(2) Execution times (i.e., time required to physically perform the action, such as handwheel a valve open, open a breaker, etc.).

(3) Communications for adequacy between the controlling location and recovery action locations for areas which involve actions.

(4) Adequate lighting (either fixed or portable) for access/egress and local lights are provided for the component to be operated.

S-3.9 1, 2 CNP calculation Probabilistic Risk Assessment (PRA)-FIRE-4.2.1.3 and Attachment G 17663-012-LAR, "Post-Fire Human Reliability Analysis" and Technical Evaluation R1900-0026-001, "Recovery Action Transition for NFPA 805" will be reviewed and updated based on the results of the field walkdowns of the recovery actions (Item S-3.8) and procedure changes (Items S-3.5, S-3.11 and.S-3.14).

Revision 5 Page S-8

Indiana Michioan Power CNP NFPA 805 Transition Renort - Attachment S Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section / Source S-3.10 1, 2 Technical documents and procedures that relate to new Fire 4.7.1, 4.7.2, 4.7.3, and Attachment W Protection (FP) design and licensing basis (e.g., Fire Protection Program Manual (FPPM), Technical Requirements Manual, Design Basis Document, maintenance and surveillance, configuration control, training and qualification guidelines, Quality Assurance Program Document (QAPD), etc.) will be revised as needed for implementation of NFPA 805.

S-3.11 1, 2 A new restoration procedure (1/2-OHP-4025-R-XX series) will be Attachment W developed to address re-powering the hydrogen igniters following a fire in Fire Areas AA40, AA43, AA46, AA47, AA48, AA50, AA51 and AA52 S-3.12 1, 2 The current transformer evaluation (Technical Evaluation 12.6) will 4.2.1.1/Attachment B be updated to address those CTs that currently have not screened out as sufficient CT data becomes available.

S-3.13 1, 2 Update the UFSAR following the guidance provided in Frequently 4.7.1/ Response to RAI-62 documented in Asked Question (FAQ) 12-0062 (ADAMS Accession No.

I&M Letter dated May 1, 2013 ML121430035).

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[-ItL*.tiI IIS IS IL Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section / Source S-3.14 1, 2 Applicable operating procedures will be revised to include the Response to RAI-61 documented in I&M treatment of Fire PRA Actions added to the 'base fire PRA model' Letter dated May 1, 2013 that mitigate 'fire induced failures - but are not associated with the NSCA success path. This includes isolation of containment purge line isolation valves, and procedure changes for Turbine Driven Auxiliary Feed Water (TDAFW) pump battery charger re-alignment to address uncertainty considerations. This includes procedure changes associated with the NSCA (item S-3.5), field verification walk-throughs (item S-3.8), hydrogen igniters (item S-3.11) and temporary ventilation (item S-3.17).

S-3.15 1,2 Revise procedure PMP-2270-WBG-001, "Welding, Burning and Response to RAI-09.01 documented in I&M Grinding Activities," and procedure 12-FPP-2270-066-011, "Fire Letter dated October 15, 2012, and Watch Activities," and conduct training on discontinuing the use of Attachment A.

(1) video cameras for fire watch and (2) use of a single fire watch for multiple hot work activities.

S-3.16 1,2 Revise Procedure PMP-2270-CCM-001, "Control of Combustible Response to RAI-10.01 documented in I&M Materials," and conduct training on the requirements of NFPA 805 Letter dated October 15, 2012, and Section 3.3.1.2(1).

Attachment A.

S-3.17 1,2 Revise Technical Evaluation R1900-0026-001, Revision 1, Response to RAI-18.01 documented in I&M "Recovery Action Transition in Support of NFPA 805," and Letter dated October 15, 2012, and applicable procedures, to establish temporary MCR ventilation for Attachment A.

Recovery Actions associated with VFDRs for AA3-004, AA3-009, AA36/42.42-026, AA57A-001 and AA57B-001. The Implementation Item will also require that training be conducted.

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Indiana Michinain Power CNP NFPA 805 Trans~ition Renon

- Attachme~nt S Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section / Source S-3.18 1,2 Verify MOV circuit changes (Table S-2, Item S-2.1) have been Response to RAI-44 documented in I&M accomplished as credited in Attachment W and verify the Letter dated October 15, 2012.

associated valves are not credited in other areas where spurious operation could occur.

S-3.19 1,2 Update the associated Fire PRA task and revise the Fire PRA Response to RAI-44 documented in I&M ignition frequency and/or fire modeling to include re-evaluation of Letter dated October 15, 2012.

procedural actions/controls associated with control of automatic CO 2 actuation in certain areas (Table S-2, item S-2.2), Transient Combustible Fee Zones (Table S-3, Item S-3.3), establishing certain areas as Hot Work Restriction Zones (Table S-3, Item S-3.4).

3-3.20 1,2 Upon completion of all Fire PRA credited implementation items in Response to RAI-44 documented in I&M Transition report Tables S-2 and S-3, verify the validity of the Letter dated October 15, 2012.

change-in-risk provided in Attachment W. This includes procedure Response to RAI-61 documented in I&M changes affecting the fire PRA (item S-3.14), electrical circuit Lette t

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in 1&M modifications (item S-3.18), fire ignition and suppression (item S-Letter dated May 1,2013.

3.19) and PRA methods (item S-3.21).

If this verification determines that the risk metrics have changed such that the RG 1.205 acceptance guidelines are not met, additional analytical efforts, and/or procedure changes, and/or plant modifications will be implemented to assure the RG 1.205 acceptance criteria are met.

S-3.21 1, 2 Conduct a Focused Scope Peer Review of the PRA upgrade ite This implementation item has been deleted rel-ated, to Redured

,issio-n Times for. cutscts containing a Test and included as a License Condition in and Maintenance event combined with a running failur*e.

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Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section / Source S-3.22 1, 2 Revise program documents and procedures, and conduct Response to RAI-60 documented in I&M associated training, as necessary to implement specific Letter dated October 15, 2012.

requirements from NFPA 805 Section 2.7.3 as described in Transition Report Section 4.7.3.

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S-3.23 1, 2 Revise Technical Evaluation 12.5 and update the following Revised response to RAI-15(b) documented documentation to reflect the revised response to RAI-15(b) in I&M Letter dated February 1, 2013.

documented in I&M Letter dated January 14, 2013:

1. Revise Unit 1 and Unit 2 250 Volt DC calculations 1-E-N-ELCP-250-001, "Unit 1 250VDC System Coordination Study,"

and 2-E-N-ELCP-250-001, "Unit 2 250VDC System Coordination Study." These will be non-technical revisions to provide clarity regarding the CNP Licensing Basis and Design Requirements (specifically, the SSCA and NEI 00-01).

2.

Revise 600 Volt AC calculation 1-E-N-PROT-BKR-007, "Unit I 600V Switchgear Breaker 11A6, 11A7, 11B3, 11C3, 11C9, 11C18, 11D9 and OB2-1 Settings." This will be a non-technical revision to provide clarity regarding the NFPA 805 requirements in the case of a potentially overloaded cable.

3. Create new Unit 1 and Unit 2 120 Volt AC calculations. A representative sample of cable data from buses identified in Technical Evaluation 12.5 was evaluated for adequate cable protection and coordination. Results of the survey revealed that all cables were acceptable with significant margin to preclude cable damage and secondary fires.

The new calculations will document the assembled data for each 120 Volt AC cable and protective device and be complete by March 28, 2013.

4. Update Technical Evaluation 12.5. This update will provide justification for resolution of conditions currently identified as deficiencies with 250 Volt DC, 600 Volt AC and 120 Volt AC calculations.

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