Text

Response to a Request for Additional Information (RAI 63) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805
	ML13262A012
Person / Time
Site:	Cook
Issue date:	09/16/2013
From:	Lies Q; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP-NRC-2013-75, TAC ME6629, TAC ME6630
	Download: ML13262A012 (42)
	v • d • e

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power I nd iana Michigan Power.com September 16, 2013 AEP-NRC-2013-75 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 63) 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 (I&M), 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).

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, 10, 13, and 17 the NRC transmitted RAIs regarding the proposed amendment. References 5, 6, 7, 11, 12, 14, 15, 16, and 18 transmitted I&M's responses to the Reference 3, 4, 8, 9, 10, 13, and 17 RAIs. By Reference 19, the NRC transmitted RAI 63 regarding a change in risk evaluation using alternative key assumptions (Reference 16), specifically the ability of proposed reactor coolant pump (RCP) seals that, when actuated, are expected to restrict reactor coolant system inventory losses to very small leakage rates during plant events that arise from the loss of all RCP seal cooling. This letter provides I&M's response to Reference 19. to this letter provides an affirmation statement. Enclosure 2 identifies documents referenced in this letter and its enclosures. Enclosure 3 is the response to RAI-63. Enclosures 4 and 5 provide revisions to Attachments M and S, respectively, of the Transition Report.

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.

U.S. Nuclear Regulatory Commission Page 2 AEP-NRC-2013-75 There are 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, Shane Lies Vice President, Indiana Michigan Power JMT/amp

Enclosures:

1. Affirmation

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

3. Donald C. Cook Nuclear Plant Response to Request for Information 63

4. Revision 6 of Attachment M, "License Condition Changes," to the Transition Report Provided in Support of Response to Request for Additional Information Concerning NFPA 805 LAR Supplement Dated September 6, 2013

5. Revision 7 of Attachment S, "Plant Modifications and Items to be Completed During Implementation," to the Transition Report, Provided in Support of Response to Request for Additional Information Concerning NFPA 805 LAR Supplement Dated September 6, 2013 c:

J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosures MDEQ - RMD/RPS NRC Resident Inspector C.D. Pederson, NRC Region III T. J. Wengert, NRC Washington, D.C.

to AEP-NRC-2013-75 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am a 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 Q.

hane Lies Vice President, Indiana Michigan Power SWORN TO AND SUBSCRIBED BEFORE ME THIS l DAY OF 2013 My Coi Notaqnx blic (

My Commission Expires Iý'-\\_.*_*3 DANIELLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 04-04-2018 Acting In the County ot*

x L.

to AEP-NRC-2013-75 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. ML11188A145.

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-2013-75 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. 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.

11. 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. ML12297A213.

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 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.

13. 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 December 13, 2012, ADAMS Accession No. ML12345A327.

14. 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.

15. 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-2013-75 Page 3

16. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "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),"

AEP-NRC-2013-17, dated May 1, 2013, ADAMS Accession No. ML13123A298.

17. E-Mail from T. J. Wengert, NRC, to H.L. Etheridge, 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 June 14, 2013.

18. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant, Units 1 and 2, Response to a 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) and the Use of Compensatory Measures Associated With the Proposed Modifications," AEP-NRC-2013-55, dated June 21, 2013, ADAMS Accession No. ML13178A209.

19. 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 September 6, 2013.

to AEP-NRC-2013-75 Donald C. Cook Nuclear Plant Response to Request for Information 63

Background

By letter dated May 1, 2013, (ADAMS Accession No. ML13123A298) Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, provided a response to RAI-61, which requested a change in risk evaluation using alternative key assumptions. Based, in part, on the results of this evaluation, the licensee proposes to install new reactor coolant pump (RCP) seals that, when actuated, are expected to restrict reactor coolant system inventory losses to very small leakage rates during plant events that arise from the loss of all RCP seal cooling. Installation of these new seals is expected to reduce risk from seal leakage and this reduction in risk caused the total change in risk from transitioning to National Fire Protection Association Standard 805 (NFPA 805) using alternative assumptions to decrease to a value below the acceptance guidelines in Regulatory Guide (RG) 1.174. The likelihood and magnitude of inventory loss from these types of RCP seals assumed by CNP was based on Revision 1 to Pressurized Water Reactor Owners Group Topical Report WCAP-171 OO-P/NP, "PRA Model for the Westinghouse Shutdown Seal."

Recently, I&M notified the Nuclear Regulatory Commission (NRC) about recent Operating Experience (OE) with the new RCP seals, which indicates that the likelihood and magnitude of inventory loss may be greater than that assumed by CNP from these types of seals.

Request for Additional Information (RAI)-63 Probabilistic Risk Assessment

1. What impact does this recent operating experience with the new RCP seals have on the reduction in risk referred to above?

Provide the revised risk estimates of the impacts, including the basis for any revised assumptions.

2. If the impact causes the total change in risk from transitioning to NFPA 805 using alternative assumptions to exceed the acceptance guidelines of RG 1.174, how does CNP plan to address this operating experience?

3. Discuss whether any currently proposed license conditions are affected by these changes.

RAI-63

Response The response to RAI-63 is written in the following three parts, corresponding to the NRC questions.

Part 1 - Risk Impact of OE.

Part 1 of the RAI response provides the revised risk estimates and basis.

Part 2 - CNP Plans to Address the RCP Seal OE. Since the risk impact in Part 1 shows the quantified risk of a conservative risk assessment to slightly exceed the acceptance guidelines of RG 1.174, this section discusses the CNP plan to address this operating experience.

to AEP-NRC-2013-75 Page 2 Part 3 - Changes to License Conditions. Based on the revised CNP plans in Part 2, this section describes the proposed changes to the license conditions associated with transition to NFPA 805.

Part 1 - Risk Impact of Recent OE.

In July 2013, OE in the nuclear industry revealed that the Westinghouse SHIELD seal could not be credited to perform as was previously documented in WCAP 17100 and the associated NRC Safety Evaluation. In order to estimate the impact of the potentially degraded RCP seals on the CNP transition risk, the aggregate sensitivity study which was developed in response to RAI-61 in April 2013 was re-quantified without crediting the RCP SHIELD seal modification.

The total core damage frequency (CDF) and total large early release frequency (LERF) results are within the 1E-4/yr risk guidelines of RG 1.174 and need no further consideration. However, the delta-CDF and delta LERF for each unit exceeds the RG 1.174 guidelines when credit for the RCP SHIELD seal modification is removed (1.7E-5/yr and 1.9E-5/yr for Unit 1 and Unit 2, respectively).

Revision 0

of calculation PRA-FIRE-17663-015-RAI-61 (CNP Fire Probabilistic Risk Assessment (PRA) Uncertainty and Sensitivity Analyses) documented the development of the aggregate sensitivity cases as requested in the CNP April 2013 response to RAI-61 (Reference AEP-NRC-2013-17).

The RAI-61 response credited the RCP SHIELD seal modification.

Revision 1 to calculation PRA-FIRE-17663-015-RAI-61 now documents the quantification of the RAI-63 response without credit for the RCP SHIELD seal, and with the addition of the plant's Supplemental Diesel Generators.

While the initial quantification of the risk impact in Part 1 shows the delta-risk limits to be exceeded, the development in Part 2 of this RAI-63 response identifies and evaluates conservatisms in the Fire PRA (FPRA).

Part 2 shows the plant meets the RG 1.174 risk acceptance guidelines when the conservatisms are reduced.

Calculation PRA-FIRE-17663-015-RAI-63 documents the sensitivity cases performed to evaluate the impact of reducing conservatisms in the FPRA.

The evaluation of conservatisms focused on delta-CDF. At each unit, the delta-LERF is less than 10% of the delta-CDF and correspondingly follows changes in delta-CDF.

Thus, if the delta CDF can be shown to be acceptable under RG 1.174, the delta-LERF is also expected to be acceptable.

Following evaluation of individual conservatisms, Part 2 concludes with the quantification of each of the transition risk metrics.

Part 2 - CNP Plans to Address the RCP Seal OE.

The RCP SHIELD seals have been installed in Unit 1, but not Unit 2. Since the reliability of the RCP SHIELD seals is currently unknown, CNP is deferring the installation at Unit 2 until the RCP seal reliability issue is corrected. CNP recognizes the benefit of improving the RCP seal performance and will work with the RCP SHIELD seal vendor as corrective actions are developed, but based on the current OE CNP withdraws the credit for the RCP SHIELD seals from the FPRA used to evaluate the risk of transition to NFPA 805. Since withdrawing credit for the RCP SHIELD seals from the FPRA results in an increase in the calculated risk which slightly to AEP-NRC-2013-75 Page 3 exceeds the RG 1.174 delta-risk acceptance guidelines, the FPRA model was reviewed for potential changes in order to meet the transition risk guidelines.

Before the aggregated sensitivity study and credit for the RCP SHIELD seals was applied in the RAI-61 response, each of the previous NFPA 805 transition sensitivity studies conducted on individual model issues resulted in a change in risk that was acceptable (References 5, 7, 11, 12, 15, and 16). When the collective set of sensitivity studies were conducted for RAI-61, it was necessary to credit the RCP SHIELD seal modification in order to ensure the risk acceptance criteria were met. Now that it appears the RCP SHIELD seals do not meet their performance criteria, it is necessary to examine conservatisms within the FPRA as part of this RAI-63 response, as described below. The evaluation of these conservatisms is described below in the following sections:

Identification of Dominant Contributors to Delta-CDF Evaluation of Conservatisms in the Dominant Contributors to Delta-CDF Qualitative Considerations Results Summary Following Removal of RCP SHIELD Seal Modification Credit and Reduction in Conservatisms to Delta-CDF Transition Risk Metrics and Conclusion 1.0 Identification of Dominant Contributors to Delta-CDF I&M reviewed the FPRA in order to identify the dominant contributors associated with the delta-risk. The top issues contributing to the delta-risk metrics are as follows:

a) Loss of RCP Seal Cooling Leading to a Demand for Operator Action to Trip the RCP.

This issue includes the timing of fire-induced failure of loss of all RCP seal cooling and the quantification of the associated Human Error Probability for failure to trip a RCP after loss of RCP seal cooling.

b) Control Room Cable Vault (CRCV) Fires. Modeling cable vault fires with limited ability to identify cable routing within the control room cable vault.

c) Spurious Actuation Failure Probabilities for CNP Double-Break Circuits. The response to RAI-61 used surrogate probabilities of 0.5 to represent many of the circuit failure probabilities due to spurious actuation. Currently, it is expected that NUREG/CR-7150 Volume 2, when published in October 2013, will provide guidance and methods which will yield lower probabilities for certain classes of components than are currently in the FPRA.

to AEP-NRC-2013-75 Page 4 2.0 Evaluation of Conservatisms in the Top Contributors to Delta-CDF 2.1 Loss of RCP Seal Cooling Leading to a Demand for Operator Action to Trip the RCP Severe fire damage can occur from fires in the CRCV and switchgear cable vaults. There are several fire scenarios in these rooms which potentially fail three important functions:

a) RCP seal cooling b) RCP trip circuits c) RCS makeup equipment that would be used to respond to a RCP seal loss of coolant accident (LOCA)

Sustained failure of RCP seal cooling and failure of RCP trip affect the size of the RCP seal LOCA. For scenarios with a small RCP seal LOCA, RCS inventory makeup can be supplied by the Chemical and Volume Control System (CVCS) cross-tie from the other unit which provides a safe shutdown path. However, the large (480 gpm/pump) RCP seal LOCA cannot be mitigated by the CVCS cross-tie. Following failure of all RCP seal cooling, the RCPs must be tripped within 13 minutes following a loss of all RCP seal cooling in order to prevent the 480 gpm/pump RCP seal LOCA to achieve a safe shutdown path.

If the RCP trip function from the Main Control Room is failed by the fire, RCP trip must be accomplished by locally opening the RCP breakers at the bus. Thus, for some fire scenarios, failure to trip the RCP within 13 minutes of loss of RCP seal cooling will lead directly to core damage with no possibility of mitigation.

These scenarios are a dominant contributor to delta risk.

The operator action to trip the RCPs following a fire-induced loss of RCP seal cooling was required for Appendix R (formerly) and is required for NFPA 805 Safe Shutdown compliance.

CNP has demonstrated sufficient timing, procedures, equipment and training to trip the RCPs within 13 minutes and thus the action satisfies the NSCA criteria. For NFPA 805, the action is also evaluated on a probabilistic basis, which measures the reliability of this manual action.

Because the time available for response is relatively short and the time required for the action uses most of the time available, the failure probability for this local manual action is relatively high. The developments of the demand for RCP trip and the associated timeline and options for operator action are driven by conservatisms in safety analysis and constraints of fire modeling.

These conservatisms are identified and evaluated in the subsequent sections.

Identification of Conservatisms:

The conservatisms and assumptions associated with the RCP trip local manual action are discussed below.

Fire Progqression Timing.

Fire modeling techniques used for CNP for the control room, CRCV, and auxiliary CRCV fires do not provide a differential time to damage for each cable in the scenario. All cables are modeled in the FPRA to be failed at the same time. In order to be conservative, this time is assumed at time zero. NUREG/CR-6850 specifies the typical growth rate for these fires is 12 minutes for a fixed ignition source and eight minutes for a transient fire source.

In a more realistic simulation for these fires, there would be a time delay between the start of the fire and fire damage of critical components.

This would reasonably allow a time window of up to 12 minutes for operator action from the main to AEP-NRC-2013-75 Page 5 control room (MCR). The MCR action to trip an RCP is much easier, faster and simpler than the local action and consequently has a significantly lower human error probability (HEP).

Time Available for Action to be Completed.

The 13 minute time window available for operator action is based on generic evaluations in WCAP-15601 and WCAP-16141 which assume a RCP seal cartridge volume of 39 gallons and a RCP seal leakage rate of three gpm.

When forced seal cooling is lost, the hot RCS water backflows through the seal cartridge and out the leakoff line. At three gpm leakoff rate, the normal water volume of the RCP seal cartridge is replaced by hotter RCS water in 13 minutes, causing the seal to heat up.

Westinghouse has developed test data for increased seal leak rates under these conditions, with a non-rotating shaft (i.e., RCP tripped). For the case of an operating RCP, test data is non-existent. The Westinghouse seal LOCA model conservatively assumes the rotating shaft case results in the maximum size LOCA at 13 minutes. The 13 minute time available window is chosen to bound all cases. Without test data or operational experience for these conditions, it is not possible to develop a realistic time estimate.

A moderate increase in the time window from 13 minutes to 20 minutes could result in an estimated HEP reduction by a factor ranging from three to ten.

Time Required to Perform the Action. There are two factors affecting the time required to perform the action. First there is a five minute time delay modeled at the start of the fire to address completion of the Reactor Trip procedure (E-0), to confirm a fire in the affected area and to transfer to the appropriate fire procedure. Second, the operator action modeling requires time to transit to the area and to don personal protective equipment (PPE), which increases the execution time (performance time) for this manual action. The execution time to complete the local RCP trip is seven minutes (which includes transit time and PPE). The total time required for the action totals 12 minutes and has been validated by field walkdown. The time required to complete the action will be further reviewed and refined as action items to update fire procedures and training are to be accomplished during the implementation phase (Transition Report Table S-3, Items S-3.5, S-3.7, and S-3.14).

Modelinq of Power to RCP Bus. This issue affects the demand probability for RCP trip given a fire in the dominant scenarios.

RCP trip is only required for scenarios where power remains on the RCP bus. If power to the RCP bus is failed by the fire, the pump will coast down and there is no need for the operator action to trip the RCP. The scenarios of concern all cause considerable cable damage to control cables for electrical busses and breakers.

The FPRA models spurious breaker operation as well as breakers failing as-is.

For scenarios where it is not possible to positively show the RCP bus has lost power due to fire damage, RCP trip is conservatively modeled as required. The fault trees in the FPRA model a failure of the breakers to remain in the desired position. Based on the latest NRC cable tests, these spurious probabilities are assigned a value of 0.5.

If the probability of the breaker spuriously opening is 0.5, it can be inferred that the probability of the breaker remaining closed is also 0.5. However, the fault trees in the FPRA in the April 2013 RAI-61 response do not include credit for complement terms. If the failure probabilities for spurious breaker actuation were in the typical range of failures (i.e., 0.1 - 0.03), the complementary term would not cause a significant variation, and could be ignored. However, when using a 0.5 for spurious operation, failure to model the complementary condition can cause significant conservatism in the calculated frequencies for these scenarios.

to AEP-NRC-2013-75 Page 6 As described above, the modeling of the complementary condition was not incorporated in the 2011 license amendment request (LAR) FPRA due to the lower spurious failure probabilities used at the time of the LAR submittal. The breaker modeling was left "as-is" in the aggregate FPRA for RAI-61, even though the RAI-61 FPRA models spurious opening of the offsite power breakers at a failure probability of 0.5. The effect of the SHIELD seals (included in RAI-61) to protect against the large seal LOCA was significant enough to obviate the need for complementary failure state modeling.

The RAI-61 FPRA models the reserve auxiliary transformer (RAT) breakers to close within 30 seconds of fire initiation prior to any fire damage. Further discussion below will show that if a time window is allowed for RAT breaker operability after reactor trip, the RAT breakers can be opened from the main control room.

Attachment S Implementation Item S-3.14 has been updated to include an activity to improve the reliability of the operator action to trip the RCPs after a fire-induced loss of RCP seal cooling. The effectiveness of any change is measured as described in Implementation Item S-3.20. Sensitivity cases were performed to determine the impact of timing assumptions and the extent these modeling techniques over-estimate the need for RCP pump trip, when using high spurious probabilities for loss of offsite power events.

Effect of Conservatisms:

Alternate timing and modeling of the demand for RCP trip can reduce the impact of the operator action associated with RCP trip as described below.

Allowance of Five Minutes of Fire Growth before Damage to RCP Trip Circuits The modeled timing of all failures at T=0 relegates the RCP trip circuits in the MCR to a failed state. Under normal conditions, the RCPs can be tripped from main control board (MCB) panel

7 or the power supplies to the RCP busses can be opened at MCB panel #2. Either action will remove power from the RCPs and allow the RCPs to trip. Under the former Appendix R rules, the procedural direction to locally open the RCP trip breakers was a confirmatory action because Appendix R guidelines assumed sufficient time was available for MCR actions before the control room circuits were damaged by the fire. Within the current framework of the CNP FPRA, all fire damage in a scenario occurs at the same time. It is not practicable to provide timing of circuit failures relative to each other. The one exception to this rule is the RAT breaker is modeled as closing to complete the fast bus transfer after reactor trip. All other circuits are modeled as failed at T=0. If it were possible to credit the control room trip circuits for the first five minutes, the HEP for RCP trip would revert to the in-control room trip value of 5E-3.

Complementary State Modeling The fault trees do not model the complementary state from the failure state. That is, when a component spurious transfer is modeled with a 0.5 probability, the configuration where the component is in the original position is now reduced to 0.5, but is not ANDED with a basic event to indicate such in the fault tree.

to AEP-NRC-2013-75 Page 7 Impact of Reducing Conservatisms Three sensitivity case studies were performed to investigate alternate modeling and timing of fire damage.

a) Sensitivity RCPI: This sensitivity assumes that for scenarios causing fire damage to the RAT breakers, they are failed open at T=0 with all other equipment failures.

This assumption removes the requirement for local RCP trip in scenarios with fire damage to the RAT breakers, because the RCP busses, and RCPs, will always be failed if the feed breaker is open.

This sensitivity reduces Unit 1 delta-CDF by 1.OE-5/yr to 7.OE-6/yr and reduces Unit 2 by 9.2E-6/yr to 9.9E-6/yr.

b) Sensitivity RCP2: Assuming the RAT breaker circuits and RCP trip circuits are operable for the first five minutes for 50% of the fire scenarios. This allows control room trip of the RCPs via normal or via alternate means. These assumptions reduce Unit 1 delta CDF by 5.0E-6/yr to 1.2E-5/yr and Unit 2 delta-CDF by 4.7E-6/yr to 1.4E-5/yr.

c) Sensitivity RCP3: Crediting the complementary failure state for spurious operation of the RAT breakers. This removes the overestimation of scenarios where RCP trip is required. This reduction is valid in fire scenarios where the offsite power breakers sustain fire damage and are postulated to transfer position with a probability of 0.5.

These assumptions reduce Unit 1 delta-CDF by 3.5E-6/yr to 1.4E-5/yr and Unit 2 delta-CDF by 3.9E-6/yr to 1.5E-5/yr..

These sensitivity calculations show that refinements in the modeling assumptions used in the timeline of critical events can significantly reduce the delta-CDF attributed to failure of the operator action to trip RCPs.

2.2 CRCV Fires The CRCV account for 21% of Unit 1 delta-CDF and 19% of the Unit 2 delta-CDF. The CRCVs are relatively, small areas underneath the MCR (essentially a crawl space with limited head room), with primary access through a normally locked hatch in the floor of the MCR. Access to the CRCV requires a fire brigade member to be present at the point of access to isolate the C02 and Halon suppression systems, and to unlock the hatch after getting concurrence from the control room operators. There are no fixed ignition sources in the cable vaults. The CRCVs are currently maintained transient combustible free and are hot work restricted areas in accordance with good housekeeping practices.

Under NFPA 805, these rooms have been established as transient combustible free and hot work restricted areas in accordance with procedures PMP-2270-CCM-001, Revision 20, and PMP-2270-WBG-001, Revision 19, respectively.

The CRCV of each unit has two gaseous suppression systems (C02 and Halon). The predicted transient ignition frequency for either CRCV using FAQ-12-0064 is 7.2E-6/yr. Three transient fires are modeled for each CRCV. These are:

to AEP-NRC-2013-75 Page 8 a) fire with success of suppression at four minutes b) fire with success of suppression at twenty minutes c) fire with failure of all fixed suppression The major conservatism in the modeling of CRCV fires is that the majority of the critical cables are "unscheduled." That is, they are not routed in trays, conduits, or raceways, but simply are routed along the floor with an unknown routing path. Since locations of the critical cables are not known, there is no definitive engineering basis to provide credit for time to damage and/or suppression, due to inability to determine a distance between the fire source and critical cables.

Consequently, the conservative assumption was made that all "unscheduled" cables are failed for all three fire scenarios. This assumption results in all three fires having essentially the same conditional core damage probabilities, because the difference in cable damage sets for the three scenarios is insignificant.

The resulting conservatism is that CNP has a CRCV with two suppression systems which are not being credited for effectiveness in the FPRA because of limitations in the cable routing information for critical cables. For sensitivity case CRCV1 the model was revised to reflect that success of suppression in four minutes will result in failure of only one train of equipment. The CRCV sensitivity case results in a reduction of the delta risk by 3.5E-6/yr for Unit 1 and 3.9E-6/yr for Unit 2.

2.3 Spurious Actuation Failure Probabilities for CNP Double-Break Circuits The CNP FPRA Model was completed in 2010, before the completion of the NRC cable fire tests in NUREG/CR-7150, Volume 1.

In the LAR submittal, CNP credited control power transformers for the reduction of spurious operation probabilities in accordance with NUREG/CR-6850.

These credits were removed in the response to RAI-61.

The RAI-61 sensitivity study conservatively modeled the following attributes:

a) Spurious valve probability of 0.5 for all valve probabilities that previously relied on a control power transformer.

b) No credit for hot short duration of direct current (DC) circuits.

c) No credit for hot short duration of alternating current (AC) circuits.

New guidance on modeling of hot shorts is expected to be published in NUREG/CR-7150 Volume 2 in October 2013.

This guidance is expected to allow alternate assumptions as follows:

1) AC motor operated valves (MOV) with a "double break circuit" design can be credited for spurious actuation at a probability of 0.4, as opposed to the 0.5 now being used in the RAI-61 response.

All MOVs in the CNP FPRA have double break control circuits.

2) Hot shorts for DC powered valves and AC powered solenoids, can be credited with self-healing in accordance with FAQ-08-0051.

A sensitivity case was developed that incorporated these expected changes. All MOVs credited with a spurious operation probability of 0.5 in the FPRA used in the RAI-61 response were to AEP-NRC-2013-75 Page 9 lowered to 0.4, because of credit for a double break control circuit design. All solenoid valves (DC and 120 VAC) whose spurious opening could be tolerated for five minutes without any deleterious effects were credited with a probability of self-healing of 0.1, in accordance with FAQ-08-0051. This sensitivity study reduced delta CDF by 1.3E-6/yr for Unit 1 and 2.OE-6/yr for Unit 2, and is labeled as sensitivity case SPUR1.

Note: The offsite power breakers of concern in RCP3 above are non-safety related components and do not have a "double break design".

Their spurious actuation probabilities were not changed.

3.0 Qualitative Considerations There are additional conservatisms present in the CNP FPRA which do not lend themselves to quantification. Examples of these are discussed below.

Use of 98th Heat Release Rate (HRR) for Maiority of Fire Scenarios:

The 98 th HRR was generally used as the mean value, which would result in an overestimation of fire size. The ASME/ANS FPRA standard requires use of two point fire modeling, but that does not necessarily mean "use of two HRRs." For risk-significant fires, two points were modeled, one being the severity factor for damage to the ignition source only, and the other for impacts beyond the source.

For the latter, the 9 8th percentile was predominately used which conservatively bounds a large range of HRRs. This method meets Category II of the ASME PRA standard for two point modeling, but also results in the use of the 98th HRR for fire damage beyond the source, which is the second point, and produces conservative damage assumptions.

Transient Fires in Combustible Free Zones:

Approximately half (53% for Unit 1 and 48% for Unit 2) of the delta risk is caused by unsuppressed transient fires in the cable vaults and switchgear rooms.

As previously discussed, these rooms are equipped with automatic gaseous suppression systems.

In addition, they are transient combustible free zones with hot work limitations in the CRCVs and switchgear cable vaults (AA48/50/51/52).

It is counter-intuitive that half the plant's delta risk will be due to transient fires in rooms where precautions have been taken to prevent fires from starting and engineering systems have been installed to prohibit and suppress fire growth. This result does not comport with reactor OE, especially considering that the same FPRA predicts turbine oil/H2 fires will contribute 11% to delta risk and high energy arcing faults will contribute 4% to delta risk.

To date, there have been no transient fires of the type and magnitude in United States' nuclear history, necessary to achieve the extent of fire damage predicted in the CNP FPRA. We believe the CNP FPRA over-predicts the extent and magnitude of transient fires, because there are no historical events upon which to develop and calibrate fire modeling guidelines, techniques and ground rules. Modeling of these fires is constrained by lack of OE upon which to calibrate data and methods, in the following analytical areas.

to AEP-NRC-2013-75 Page 10 a) Calculation of fire initiating event frequency (IEF) in a combustible free zone.

b) Characterization of the fire with respect to the combustible type, combustible content, combustible configuration.

c) Characterization of the ignition source as spontaneous or human induced.

d) Heat release rate and fire growth rate.

CNP has applied the NUREG/CR-6850 guidelines in a conservative manner which leads to this result. It is difficult to call these methods "conservative" because there are no other methods available for comparison.

However, CNP believes transient fire modeling is conservative by nature of the absence of OE for such fires with which to calibrate modeling methods and data.

4.0 Results Summary following Removal of RCP SHIELD Seal Modification Credit and Reduction in Conservatisms to Delta-CDF This evaluation has shown several conservatisms which lead to over-estimation of the delta CDF for CNP. Five sensitivity studies were developed to investigate the extent to which these conservatisms over-estimate the risk. Each sensitivity study was evaluated by itself, with the results presented in Table 1. Following evaluation of individual sensitivity cases an updated, aggregate FPRA was developed in order to calculate the transition risk metrics for RAI-63.

Individual sensitivity studies were performed on the RCP trip requirement and the associated HEP for the RCP trip. Additionally, individual sensitivity studies were performed to investigate the lack of cable vault cable routing and to investigate the expected hot short probabilities to be published in NUREG/CR-7150, Volume 2 as summarized below.

Three sensitivities were performed to investigate the impact of removing or mitigating the requirement for local RCP trip in 13 minutes.

o The first sensitivity on RCP trip (RCP1) removes the RCP trip requirement. This shows the maximum bounding case for removal of the requirement.

o The second sensitivity on RCP trip (RCP2) models that 50% of all fires do not damage control cables until five minutes after fire ignition. This is reasonable when compared to the 12 minute growth curve for fixed sources and the eight minute growth curve for transient sources.

This sensitivity was used in the updated, aggregate FPRA for RAI-63.

o The third sensitivity on RCP trip (RCP3) provides credit for the complementary probabilities of the success state, which are not modeled in the fire fault trees.

This does not remove the requirement for RCP trip, but improves the quantitative accuracy of the fault tree techniques. This study was also used in the updated, aggregate FPRA for RAI-63.

A fourth sensitivity (CRVC1) models credit for the suppression systems in the CRCV.

Due to the lack of known cable routing in the CRCV, the FPRA associated with the 2011 LAR did not credit the installed suppression systems. This is also a reasonable alternative and was used in the updated, aggregate FPRA for RAI-63.

to AEP-NRC-2013-75 Page 11 A fifth sensitivity study models credit for the yet to be published guidance for hot short probabilities in NUREG/CR-7150, Volume 2.

This sensitivity represents a reasonable set of alternative data and is also used in the updated, aggregate FPRA for RAI-63.

The potential risk reduction in delta-CDF is shown in Table 1 below when alternate methods and data associated with each conservatism are evaluated individually.

An updated, aggregate FPRA was developed for RAI-63 to find the combined impact of removing the identified conservatisms as the values presented in Table 1, below, cannot be simply summed. The updated, aggregate FPRA for RAI-63 consists of the aggregate FPRA model originally developed for RAI-61 with the RCP SHIELD seal modification removed, offset by the incorporation of the Supplemental Diesel Generators and reduction of conservatisms.

Table 1: Potential Reduction in Delta CDF due to Reducing Modeling Conservatisms Sensitivity Unit I Delta-CDF Unit 2 Delta-CDF (Part 1 - RAI-61 with (Part 1 - RAI-61 with RCP SHIELD seal RCP SHIELD seal credit removed) credit removed) 1.70E-51yr 1.91 E-51yr Individual Sensitivity Case Described in Reduction in Reduction in Section Unit 1 Delta-CDF Unit 2 Delta-CDF RCP1) Eliminate requirement for 2.1 case a 1.OE-5/yr 9.2E-6/yr local RCP trip if RAT breakers failed open at T=0.

RCP2) Assume some MCR 2.1 case b 5.OE-6/yr 4.7E-6/yr circuits operable for 5 minutes for 50% of fires RCP3) Credit for complementary 2.1 case b 3.5E-6/yr 3.9E-6/yr failure state of electric power distribution breakers CRCV1) Suppression System 2.2 3.5E-6/yr 3.9E-6/yr effective for CRCV fires SPUR1) Expected Guidance for 2.3 1.3E-6/yr 2.OE-6/yr Hot Short Probabilities Aggregate Sensitivity Case Unit 1 Total Delta-CDF Unit 2 Total Delta-CDF Effects of RCP2, RCP3, 7.45E-6/yr 9.17E-6/yr CRCV1, and SPUR1 considered simultaneously to AEP-NRC-2013-75 Page 12 5.0 Transition Risk Metrics and Conclusion This evaluation discussed several conservatisms which lead to over-estimation of the delta CDF for CNP. The risk reduction possible with alternate methods and data associated with each conservatism, evaluated individually, is shown above in Table 1.

The NFPA 805 transition risk of the CNP units is adequately represented by the aggregate sensitivity study shown in Table 1 above. This study shows an acceptable risk for transition to NFPA 805 for each unit. Table 2 below shows the NFPA 805 transition risk metrics, considering reduction in FPRA modeling conservatism as previously discussed, which partially offset the impact of removing credit for the RCP SHIELD seal modification.

Table 2: Risk Metrics of the CNP FPRA with RCP SHIELD Seal Credit Removed and Conservatisms Refined PARAMETER UNIT I UNIT 2 RG 1.174 Criteria TOTAL FIRE CDF 3.13E-5/yr 2.64E-5/yr Total CDF for all PRA Initiating Events (per year)

<1 E-4/yr TOTAL FIRE LERE 2.61E-6/yr 2.1OE-6/yr Total LERF for all PRA Initiating (per year)

Events <1 E-5/yr DELTA ODE (per year) 7.45E-6/yr 9.17E-6/yr Delta CDF from Fire <1E-5/yr DELTA LERF (per year) 6.12E-7/yr 7.29E-7/yr Delta LERF from Fire <1E-6/yr Table 3 shows the total CDF and total LERF, from all hazard groups, following transition to NFPA 805.

to AEP-NRC-2013-75 Page 13 Table 3: CDF and LERF at CNP After Transition to NFPA-805 HAZARD UNIT I UNIT 2 GROUP CDF LERF CDF LERF (per year)

(per year)

Internal Events including 1.33E-5 2.70E-6 1.32E-5 2.70E-6 Internal Flood Internal Fire 3.13E-5 2.61 E-6 2.64E-5 2.1OE-6 Seismic 3.17E-6 9.82E-7 3.17E-6 9.82E-7 Other External negligible negligible negligible negligible Events TOTAL 4.78E-5 6.29E-6 4.28E-5 5.78E-6 Table 2 and Table 3 demonstrate that the risk metrics associated with transition to CNP Units 1 and 2 is acceptable and is within the guidelines of RG 1.174.

NFPA 805 at Part 3 - Chanqes to License Conditions.

Based on the revised CNP plans in Part 2, this section describes the proposed changes to the implementation items and the license conditions associated with the NFPA 805 transition. The changes to the License Conditions are delineated by revision marks and include the following:

Deletion of Item 5 in the Transition License Conditions, new paragraph numbering and formatting to simplify referencing within the License Condition, and correction of 3 typographical errors. Proposed changes to the license condition are provided in Enclosure 4 to this letter as Attachment M of the Transition Report.

to AEP-NRC-2013-75 Revision 6 of Attachment M, "License Condition Changes," to the Transition Report Provided in Support of Response to Request for Additional Information Concerning NFPA 805 LAR Supplement Dated September 6, 2013 Changes are indicated by revision bars in the right margin.

Indiana Michiaan Power CNP NFPA 805 Transition Reoort - Attachment M M. License Condition Changes 8 Pages Revision 6 Page M-1 Revision 6 Page M-1

Indiana Michiaan Power CNP NFPA 805 Transition ReDort - Attachment M 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.

The following is the proposed License Condition for paragraph 2.C.(4) for Unit 1:

Fire Protection Program Indiana Michigan Power Company 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's amendment request dated July 1, 2011, as supplemented by letters dated September 2, 2011, April 27, 2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013, February 1, 2013, May 1, 2013, June 21, 2013, and September 16, 2013, and as approved in the Safety Evaluation dated

,2013.

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.

(a) 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 (FPRA) 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.

1. 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.

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Indiana Michiaan Power CNP NFPA 805 Transition Rer)ort - Attachment M

2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than lxl07/Iyear (yr) for CDF and less than lx10 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.

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

1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program and Design Elements 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 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).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

Revision 6 Page M-3

Indiana Michiaan Power CNP NFPA 805 Transition Report - Attachment M

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 dated

, 2013, 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.

(c) Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2.C.(4)(c)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.C.(4)(b)2. above.

2. The licensee shall implement the modifications to its facility, as described in, Attachment S, Table S-2, "Plant Modifications Committed," of I&M letter AEP-NRC-2013-75, dated September 16, 2013, to complete the transition to full compliance with 10 CFR 50.48(c) by _,

2014. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall implement the items listed in Enclosure 5, Attachment S, Table S-3, "Implementation Items," of I&M letter AEP-NRC-2013-75, dated September 16, 2013, by _,2014.

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 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.

Revision 6 Page M-4 Revision 6 Page M-4

Indiana Michiaan Power CNP NFPA 805 Transition Report - Attachment M The following is the proposed License Condition for paragraph 2.C.(3)(o) for Unit 2:

Fire Protection Program Indiana Michigan Power Company 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's amendment request dated July 1, 2011, as supplemented by letters dated September 2, 2011, April 27, 2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013, February 1, 2013, May 1, 2013, June 21, 2013, and September 16, 2013, and as approved in the Safety Evaluation dated

,2013.

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.

I.

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 (FPRA) 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.

1. 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.

2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than lxl0 7/year (yr) for CDF and less than lx10 -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.

Revision 6 Page M-5

CNP NFPA 805 Transition Report - Attachment M Indiana ichinan Po er 1I.

Other Changes that May Be Made Without Prior NRC Approval

1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program and Design Elements 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 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).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

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 dated

, 2013, to determine that certain fire protection program changes meet the minimal criterion. The Revision 6 Page M-6

Indiana Michiqan Power CNP NFPA 805 Transition Reoort - Attachment M licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

I1l.

Transition License Conditions

1. Before achieving full compliance with 10 CFR 50.48(c), as specified by 2.C.(3)(o)(111)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.C.(3)(o)(11)2. above.

2. The licensee shall implement the modifications to its facility, as described in, Attachment S, Table S-2, "Plant Modifications Committed," of I&M letter AEP-NRC-2013-75, dated September 16, 2013, to complete the transition to full compliance with 10 CFR 50.48(c) by

, 2014. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3. The licensee shall implement the items listed in Enclosure 5, Attachment S, Table S-3, "Implementation Items," of I&M letter AEP-NRC-2013-75, dated September 16, 2013, by

,2014.

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 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 Revision 6 Page M-7

CNP NFPA 805 Transition Report - Attachment M Inian Michinan PowerI~

IIVV 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.

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 6 Page M-8 to AEP-NRC-2013-75 Revision 7 of Attachment S, "Plant Modifications and Items to be Completed During Implementation," to the Transition Report Provided in Support of Response to Request for Additional Information Concerning NFPA 805 LAR Supplement Dated September 6, 2013.

Changes are indicated by revision bars in the right margin.

Indiana Michiclan Power CNP NFPA 805 Transition Report - Attachment S S. Plant Modifications and Items to be Completed during Implementation 13 Pages Revision 7 Page S-1

Indiana Michician Power CNP NFPA 805 Transition Report - Attachment S 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 7 Page S-2

Ind~iana~ Mit-hirn~n Poweur CNP NFPA 805.' Tramzitinn Penor - Attac'hment R' 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.

Compensatory measure:

alternate source in Fire Areas None; P edures AA14, AA23, AA39B and None; Procedures AA45B were not accurately updated.

identified.

Revision 7 Page S-3

Inrlinnn Akf-hirynn Pnwar Indi~n~Ahirhi~n Pniyt~r (AIP AIFIPA RflTrm nczitinn Penort -AttrarhmentS.

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)-CM250 stable condition.

absence or bypass of thermal overload Compensatory measure for contacts, results in NFPA 805: Appropriate continuous energization Compensatory Measures will be of the valve motor and established when the NFPA 805 potential mechanical fire protection program is damage to the valve such approved by the NRC and that manual operation via remain in place until this handwheel would be modification is complete.

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

Revision 7 Page S-4 Revision 7 Page S-4

Indiana Michiaan Power CNP NFPA 805 Transition ReDort - Attachment 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 C02 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 mitigation b) AA43 NFPA 805: Appropriate Compensatory Measures will be established when the NFPA 805 fire protection program is approved by the NRC and remain in place until this modification is complete.

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

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

Indiana Michiqan 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 7 Page S-6

Indiana Michiaan Power CNP NFPA 805 Transition Report - 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 7 Page S-7 Revision 7 Page S-7

Indiana Michidan Power CNP NFPA 805 Transition Renort - Attachment S Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section I 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 7 Page 8-8 Revision 7 Page S-8

Indi-nnn Mrhirynn PniAlonr InrIi.A*AhUiwrhe.*,l.,*n P~ic vi\\.P

,PA 805 Tra nsition Reor

- Attac ohment S Attachment S - Table S-3 Implementation Items Item Unit Description LAR Section I 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).

Revision 7 Page S-9 Revision 7 Page S-9

OAIP AI~PA Rfl~ Tr~ncitiAn P,~nnrt - A r~hm~nt.'~

rNp FPA qnli Trnnczitinn Rnnnrt - Aff-grhment S Inrlinnn mighirynn Pnmlonr 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 Response to RAI-63 documented in I&M Auxiliary Feed Water (TDAFW) pump battery charger re-alignment Letter dated September 16, 2013 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). Additionally this includes changes associated with tripping the Reactor Coolant Pumps after a fire-induced loss of RCP seal cooling.

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.

Revision 7 Page S-1O Revision 7 Page S-10

Indiana Michioan Power CNP NFPA 805 Transition ReDort - Attachment 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 C02 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).

S-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 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).

Response to RAI-63 documented in A&M If this verification determines that the risk metrics have changed Letter dated September 16, 2013 such that the risk metrics from the September 2013 RAI response (AEP-NRC-2013-75) are exceeded, 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 Condut a Focused Scope Pee. r Revew of the PRA upgrade item This implementation item has been deleted related to Reduced Mission Times for P-t* containiRg a Test and included as a License Condition in and Maintena*e event c.robined with a

.running failure.

Attachment M Revision 7 Page S-lI Revision 7 Page S-11

Inrliana Mirhican Power C NP NFPA B05 Trnnsifihn Ppnnrt - A ffnrhmpnt R 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.

Revision 7 Page S-12

Indiana Michiaan Power Indiana~~~~~ MihnPwrCP NFPA 805 Transqition Renrt - Attachment S~

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 s&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 1 600V Switchgear Breaker 11A6, 11A7. 11B3, 11C3, 11C9, 11C18, 11D9 and 012-1 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.

Revision 7 Page S-13 Revision 7 Page S-13

v t e Letter Sequence Response to RAI
TAC:ME6629, (Approved, Closed) TAC:ME6630, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement Administration Questions 14 August 2012 27 January 2012 27 January 2012 11 October 2012 13 December 2012 1 February 2013 6 September 2013 Answers 27 April 2012 9 August 2012 1 May 2013 21 June 2013 16 September 2013 Results Approval Other: ML11255A022, ML113120496, ML113120500, ML113130249, ML113130259, ML113130260, ML113130261, ML113130262, ML12285A179, ML13085A184, ML13085A190
MONTHYEAR2012-08-14 [Table View]

AEP-NRC-2013-75, Response to a Request for Additional Information (RAI 63) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805

Text

Enclosures:

References:

RAI-62

Background

RAI-63

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AEP-NRC-2013-75, Response to a Request for Additional Information (RAI 63) Regarding the Application for Amendment to Transition the Fire Protection Program to National Fire Protection Association Standard 805

Text

Enclosures:

References:

RAI-62

Background

RAI-63

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