ML13028A113

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

ML13028A113
Person / Time
Site:	Cook
Issue date:	01/14/2013
From:	Lies Q Indiana & Michigan Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2013-1, TAC ME6629, TAC ME6630
Download: ML13028A113 (19)

Text

z INDIANA MICHIGAN Indiana Michigan Power POWER8 One Cook Place Bridgman, MI 49106 A unit of American Electric Power Indiana Michiga nPower.com January 14, 2013 AEP-NRC-2013-1 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 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)

This letter provides Indiana Michigan Power Company's (l&M's) response to a third U. S. Nuclear Regulatory Commission (NRC) request for additional information (RAI) regarding a proposed license amendment to transition the Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, to a new fire protection program based on National Fire Protection Association (NFPA) Standard 805.

Documents referenced in this letter are identified in Enclosure 2.

By Reference 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 Standard 805, in accordance with 10 CFR 50.48(a) and (c). Reference 1, Enclosure 2, hereafter referred to as the Transition Report, provided information associated with the CNP transition to NFPA 805.

By References 3 and 4, the NRC transmitted the first round of RAIs regarding the proposed amendment. References 5, 6, and 7 transmitted I&M's responses to the Reference 3 and 4 RAIs.

By Reference 8, the NRC transmitted a second round of RAIs. Reference 9 and 10 provided I&M's responses to the Reference 8 second round RAIs. By Reference 11, the NRC transmitted three third round RAIs (RAI-20.01, RAI-30.01, and RAI-34.01). This letter provides I&M's responses to the three Reference 11 third round RAIs. to this letter provides an affirmation statement. identifies documents referenced in this letter and its enclosures. provides I&M's responses to the Reference 11 third round RAls.

U. S. Nuclear Regulatory Commission AEP-NRC-2013-1 Page 2 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.

This letter contains no new or modified regulatory commitments.

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

Sincerely, QAn~ton S. Lies Engineering Vice President JMT/

Enclosures:

1. Affirmation

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

3. Responses to Third Round Information Requests RAI-20.01, RAI-30.01, and RAI-34.01 Re.

NFPA 805 Transition 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 to AEP-NRC-2013-1 AFFIRMATION I, Quinton S. Lies, being duly sworn, state that I am Engineering 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 uinton S. Lies Engineering Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAY OF 2013 Notar7:* ublic

• My Commission Expires 4N-c.

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")\\m DANIELLE BURGOYNE Notary Public. State of Michigan County of Berrien My Commission Expires 04-,4-2018 Acting In the County of c to AEP-NRC-2013-1 Identification of Documents Referenced in this Letter and Its Enclosures Reference

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. ML1 11 188A145.

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. MLA-2132A390.

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 TransitioKhthe 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-2013-1 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 October 11, 2012, ADAMS Accession No. ML12276A344.

9. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Response to Second 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-2012-92, dated October 15, 2012, ADAMS Accession No. ML122970218.

10. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "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.

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 December 13, 2012, ADAMS Accession No. ML12345A327.

to AEP-NRC-2013-1 Responses to Third Round Information Requests RAI-20.01, RAI-30.01, and RAI-34.01 Re.

NFPA 805 Transition Documents referenced in this enclosure are identified in Enclosure 2 to this letter.

This enclosure provides Indiana Michigan Power Company's (l&M's) response to the third round U. S. Nuclear Regulatory Commission (NRC) request for additional information (RAI) transmitted by Reference 11 regarding the Donald C. Cook Nuclear Plant (CNP) transition to National Fire Protection Association Standard 805 (NFPA 805).

RAI 20.01 Probabilistic Risk Assessment In your letterý dated August 9, 2012 and the enclosure, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12242A246) and the non-public enclosure, you responded to RAI-20, Probabilistic Risk Assessment (PRA), Intemal Events PRA (IEPRA),

explaining that a gap assessment was performed on the differences in supporting requirements (SRs) between PRA Standard RA-Sa-2009, as clarified by Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Result for Risk-Informed Activities", Rev, 2, and RA-Sa-2003, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (which was initially used to review the internal events PRA). As part of the assessment, gap findings were generated and a sensitivity study was performed to address a subset of the issues. Clarify the following gap finding:

a.

A finding against SR LE-B3, which requires using engineering analysis to support PRA modeling, identifies two large early release frequency (LERF) modeling assumptions that were not justified with engineering assessment but have beneficial impacts on the LERF estimate. The disposition for this gap finding states that these assumptions are expected to be confirmed by analysis.

The assumptions are not identified. Identify the assumptions made, and explain why they are expected to be confirmed by analysis.

b.

A finding against SR LE-A4, which requires dependencies between the Level 1 and Level 2 PRA to be properly treated, identifies that in some cases, assumptions were made that may yield non-conservative results. The disposition to this finding states that the licensee's use of NUREG/CR-6595, "An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events," produces a bounding model but also states that a sensitivity study examines the impact of this issue.

The finding against LE-A4 is not identified in the sensitivity study identified for this RAI on page 22 of the response. Clarify how the finding against LE-A4 is being dispositioned. Include in this description how dependencies between Level 1 and Level 2 are treated and clarify how SR LE-A4 is met.

to AEP-NRC-2013-1 Page 2 Response to RAI 20.01

a. The finding noted against SR LE-B3 identifies two assumptions that do not provide references to engineering analyses. These two assumptions are quoted and discussed separately below.

The first assumption identified in the gap assessment discussion of SR LE-B3 is Assumption 2 from PRA-NB-LE, Section 2. The entire assumption text is provided below:

In the medium and large Loss of Coolant Accident (LOCA) cases where an accumulator fails, it is assumed that injection of only two accumulators is insufficient to reflood/refill the core before damage occurs. In the CNP PRA model, core damage is defined in the event tree notebook (Reference 1) as occurring "... if the maximum core temperature exceeds 1400F."

The notebook goes on to state: "This temperature is sufficiently below the temperature where core damage will occur, but is above the point where sufficient core uncovery has occurred that little time is available for recovery action." In the case of LOCA with accumulator failure; however, the recovery of core cooling is purely automatic, i.e., the auto-initiation of all six of the Emergency Core Cooling System (ECCS) pumps.

It is assumed that this will happen quickly enough to assure that significant core geometry degradation has not occurred since an injection flow of only 500 gallons per minute (gpm)

(or 2/4 High High Safety Injection (HHSI) pumps or 1 Low High Safety Injection (LHSI) pump using safety analysis values) will steam cool the core as quickly as 200 seconds after the break. Even if the peak core temperature rapidly increases during this time frame, core damage (such as clad melting) will occur in the upper regions of the core first and some time will elapse before candling occurs.

The foregoing estimate is conservative because we should have two accumulators injecting (in most cases, since the success criteria is 3 out of 4), which should cause the vessel to refill rapidly with some pumped injection success.

Potentially, an argument could even be made that such sequences should not even be core damage.

Based on the above, it is assumed that these sequences (accumulator failure but success in low pressure or high pressure injection) will result in the core damage being arrested before vessel breach.

This assumption is expected to be confirmed based on its inherent reasonableness to experienced thermal-hydraulic analysts. Even in the case that all accumulators fail to inject, at least one of the low head ECCS pumps would provide ECCS injection flow within one minute of the break initiator.

Although core overheating Would occur during this interval, core geometry would be maintained and introduction of ECCS injection would prevent further core degradation.

Further, a Large, Early Release due to hydrogen deflagration is considered by the LERF model even though the extent of core damage in these scenarios is clearly very limited. Accordingly, no further action is believed required to address this finding.

The second assumption identified in the gap assessment discussion of SR LE-B3 is the first paragraph from PRA-NB-LE, Section 4.1. The entire assumption text is provided below:

Some of the CNP core damage sequences do not have core damage until after the containment has failed due to gradual overpressurization (i.e., essentially a static loading resulting from a loss of containment heat removal). However, even though core damage to AEP-NRC-2013-1 Page 3 and the vessel failure will occur inside a failed containment, these sequences will not result in a LERF.

From Reference 3, because of the CNP containment design, the gradual overpressurization will result in a failure at the basemat junction with the containment wall.

This failure will be small in nature, and the liner would be able to accommodate large differential displacements of the shell and mat after the formation of the crack without gross rupture of the liner. The small, localized tears in the liner would result in a leakage path small enough to not be considered a Large, Early Release, yet they are large enough to limit any additional buildup of pressure in containment in a manner analogous to a containment vent.

This assumption is reasonable for the scenarios of concern. By their nature, these scenarios are slowly developing since containment over-pressurization failure will not occur for several hours following ice melt-out. Since the containment failure mechanism is static over-pressurization, the containment failure mode described is qualitatively correct, but no exact estimate of containment failure area was determined in the containment structural analysis cited. However, the containment failure area required to vent the steam produced by core decay heat is reasonably expected to be very much smaller than the containment failure area caused by a sudden, massive over-pressurization event such as would occur due to hydrogen deflagration/detonation or following high pressure melt ejection.

The discussion above focused on the failure size to conclude that these sequences would not lead to Large, Early Release, but it was mute on the timing of the event, which is also an important consideration. The event timing is dependent on the specific details of the sequences of interest.

To determine the sequences of interest, the results of the IEPRA model were reviewed to identify whether any loss of containment heat removal scenarios contributed to the Core Damage Frequency (CDF).

The resulting ten sequences include one large LOCA scenario and nine variations of small LOCA scenarios (either reactor coolant system (RCS) breaks or reactor coolant pump (RCP) seal leaks/LOCAs). Although non-LOCA event trees include loss of containment heat removal sequences, none of these sequences were numerically significant in the base IEPRA model. Note that all of the sequences of interest have successful operation of ECCS injection and recirculation flow, but that there is no heat removed by the associated recirculation heat exchangers.

For the scenarios of interest, an unisolable RCS breach occurs either as the initiating event or relatively shortly after sequence initiation. The unisolable RCS leak will cause activation of the Emergency Response Organization (ERO). Since core damage occurs after containment failure for these scenarios, the ERO will be activated and be fully operational for many hours prior to any significant fission product release to the environment. Given the loss of RCS boundary, the ERO will declare a Site Area Emergency when the containment pressure exceeds its design value. This declaration will initiate off-site responses, including state-directed evacuation of hospitals, schools, and nursing homes within the evacuation zone. Given the continued loss of containment heat removal and the prospects for recovery, the ERO could make an anticipatory General Emergency declaration, and initiate a Protective Action Recommendation (PAR) to evacuate the general population prior to the actual occurrence of each condition required for the General Emergency declaration. However, an early evacuation is not certain, so the continuing sequence development must be further considered.

For the sequences of interest, the failure of containment provides a vent path for steam generated by the continuing recirculation flow through the core. The containment pressure will decrease and to AEP-NRC-2013-1 Page 4 so will the containment water inventory. The failure mechanism for the ECCS pumps will either be the high recirculation sump water temperature or the decrease of sump level below the minimum allowable value. Following loss of the ECCS pumps, the core will begin to overheat and the lack of core cooling will cause the General Emergency conditions to be satisfied. Assuming that the PAR is general population evacuation, the sirens will sound and initiate the evacuation within one hour of these conditions.

Since the core overheating and melt process is relatively slow, the effective implementation of the ERO as described above protects the public from early health effects. As a result, the classification of these scenarios as core damage but not Large, Early Release is considered justifiable.

The Fire PRA event trees were also reviewed to determine the applicability of this finding. Similar functional requirements were included in the Fire PRA. Some of these Fire PRA sequences also involve unisolable RCS breaches so the discussion above is directly applicable to these Fire PRA sequences.

The remaining Fire PRA sequences involve losses of steam generator inventory makeup and subsequent bleed and feed of the RCS. For these scenarios, the loss of heat sink condition will cause activation of the ERO; however, the remainder of the discussion provided above remains applicable. Based on the same reasoning as provided above, the classification of these scenarios as core damage but not Large, Early Release is also considered justifiable,

b. The finding noted against SR LE-A4 is related to treatment of dependencies between the Level 1 and Level 2 models. The statement from the gap assessment discussion of SR LE-A4 is as follows:

"However, in some cases, assumptions are utilized. For example, if Auxiliary Feed Water (AFW) is not questioned for a sequence the assumption is made that AFW is successful. For applications, this may yield non-conservative LERF results."

The event tree structures used for the IEPRA model and the Fire PRA model characterize sequence progression following an initiating event by addressing the automatic plant response and the proceduralized operator response to bring the plant to a safe, stable state. These event trees include automatic and manual operation of systems used to control the containment atmosphere within acceptable limits. Three possible sequence outcomes are represented on the Level 1 event trees: no core damage ("OK"); core damage, but no large, early release of radioactivity from containment ("CD"); and core damage with a large, early release of radioactivity from containment

("LER").

Given this approach, dependencies that LERF sequences have on related Level 1 sequences are completely represented.

However, as is often the case in developing a Level 1 PRA model, some sequences are readily seen to be very unlikely and/or certain characteristics obviate the efficacy of some other front-line mitigating system. For example, any small-break LOCA scenario in which all injection has failed will experience core damage, regardless of the status of steam generator inventory or containment heat removal. The event trees have consistently determined the status of containment-related systems even for sequences such as these. The event trees do not always determine the status of Level 1 mitigating systems, such as AFW, or operator actions, such as the procedurally-directed RCS cooldown to cold shutdown for small-break LOCAs, when their status does not influence the progression to core damage.

to AEP-NRC-2013-1 Page 5 This finding was investigated by reviewing the IEPRA event trees. A number of instances were found where the top event for AFW is not questioned. These occurrences fall into one of the following categories:

Unisolated Interfacing System LOCA (ISLOCA) sequences are assumed to be large, early releases regardless of subsequent mitigation response.

For these cases, a Conditional Large Early Release Probability (CLERP) of 1.0 is applied because containment bypass flow paths exist and AFW status cannot affect fission product release directly to the environment.

Station Blackout scenarios on transfer event tress where the sequence has already had AFW success. In these cases, the CLERP applied corresponds to an AFW success case because a successful AFW status has already been established.

" Sequences where a failure to trip the reactor follows any other initiating event than a transient initiator.

In these cases, there are no engineering analyses available to establish time frames for mitigation response, so proceeding directly to core damage or large, early release outcomes simplifies the event tree logic.

In these cases, the CLERP applied corresponds to an AFW failure case. This is considered conservative since the AFW failure probability is extremely small and AFW is not failed by the initiating event.

Sequences where Anticipated Transient without SCRAM Mitigation Signal Actuation Circuitry (AMSAC) fails when demanded following a failure of reactor trip.

In these cases, proceeding directly to core damage or large, early release outcomes simplifies the event tree logic. In these cases, the CLERP applied corresponds to an AFW failure case.

Loss of Offsite Power sequences where a Pressurizer power operated relief valve (PORV) sticks open following the reactor trip caused by the loss of offsite power. In these cases, proceeding directly to core damage or large, early release outcomes simplifies the event tree logic. In these cases, the CLERP applied corresponds to an AFW failure case. This is considered conservative since at least the Turbine Driven Auxiliary Feedwater Pump (TDAFP) is available.

" Sequences where a Pressurizer PORV sticks open following the reactor trip caused by a support system initiator. In these cases, proceeding directly to core damage or large, early release outcomes simplifies the event tree logic.

In these cases, the CLERP applied corresponds to an AFW failure case. This is considered conservative since the AFW failure probability is extremely small and AFW is not failed by the initiating event.

Small LOCA sequences where high pressure injection has failed and a rapid RCS cooldown followed by low pressure injection is required (Top Event OLI). AFW success is required as part of the logic for OLI, and AFW failure will cause OLI failure.

In the IEPRA, OLI is dominated by the operator failure to perform the required action and AFW failure cutsets are a minute contributor to the failure of the Top Event. In these cases, the CLERP applied corresponds to an AFW success case. This is considered a very small non-conservatism since AFW failure is such a small contributor.

Medium LOCA sequences where accumulators have failed and a second failure occurs, which involves either high pressure injection or recirculation, or containment spray injection.

In these cases, proceeding directly to core damage or large, early release outcomes simplifies the event tree logic.

In these cases, the CLERP applied corresponds to an AFW success case. This is considered non-conservative, however, to AEP-NRC-2013-1 Page 6 these scenarios are very low probability since they involve Medium LOCA initiators and failures of two independent systems, and, as a result, the non-conservatism is estimated to be very small.

During the IEPRA event tree review, a number of instances were found where the top event for the operator action to cool down the RCS to cold shutdown following a small-break LOCA initiator or RCP seal LOCA is not questioned.

These omissions always occur following AFW success, because no subsequent credit is taken for low pressure injection sources. In these cases, omitting the logic associated with asking the RCS cool down Top Event simplifies the event tree logic and does not affect assumed sequence progressions. In these cases, the CLERP applied corresponds to an RCS cool down failure case (i.e., high RCS pressure). This is considered conservative since the failure probability for RCS cool down is small and it is not failed by the initiating event.

In the Fire PRA model, separate event trees were developed and tailored specifically to address fire-related scenarios. However, the Fire PRA model includes the following instances where the top event for AFW is not questioned:

Sequences in which an ISLOCA pathway is caused by the fire initiator. For these cases, a CLERP of 1.0 is applied. For the ISLOCA scenarios, containment bypass flow paths exist and AFW status cannot affect fission product release directly to the environment.

Sequences where a failure to trip the reactor follows the fire initiator. In these cases, there are no engineering analyses available to establish time frames for mitigation response, so proceeding directly to a core damage outcome simplifies the event tree logic. These cases are treated as only core damage since they are very low probability and some mitigating systems would be available.

Station Blackout sequences following a fire.

In these cases, proceeding directly to core damage or large, early release outcomes simplifies the event tree logic. In these cases, the CLERP applied corresponds to an AFW failure case. This is considered conservative since the TDAFP is available.

Sequences where the operators fail to prevent Pressurizer overfill following a fire initiator and both the Pressurizer and Safety PORVs fail to open.

In these cases, there are no engineering analyses available to establish time frames for mitigation response, so proceeding directly to a core damage outcome simplifies the event tree logic. These cases are treated as only core damage since they are very low probability and some mitigating systems would be available.

Sequences where the fire initiator causes a small LOCA and high pressure injection fails.

These scenarios require a rapid RCS cooldown followed by low pressure injection (Top Event OLI). AFW success is required as part of the logic for OLI, and AFW failure will cause OLI failure. In the Fire PRA, OLI is dominated by the operator failure to perform the required action and AFW failure cutsets are a very small contributor to the failure of the Top Event.

In these cases, the CLERP applied corresponds to an AFW success case.

This is considered a very small non-conservatism since AFW failure is such a small contributor.

In addition, the Fire PRA event trees also include instances where the top event for the operator action to cooldown the RCS to cold shutdown following a small-break LOCA (OLI) is not questioned. These omissions always occur following AFW success, because no subsequent credit is taken for low pressure injection sources.

In these cases, omitting the logic associated with asking the OLI Top Event simplifies the event tree logic and does not affect assumed sequence to AEP-NRC-2013-1 Page 7 progressions.

In these cases, the CLERP applied corresponds to an OLI failure case (i.e., high RCS pressure). This is considered conservative since the OLI failure probability is small and OLI is not failed by the initiating event.

The majority of situations that could be affected by including explicit consideration of AFW status in an event tree involve using an over-estimate for the value of CLERP applied to the affected sequences; these situations do not yield non-conservative results.

There are some situations included in the event trees which are non-conservative, but these cause very small numerical effects because the AFW system is highly reliable (i-e., three safety-related trains and a hard-piped inter-unit cross-tie).

In addition, all of the situations that could be affected by including explicit consideration of RCS cool down in an event tree involve using an over-estimate for the value of CLERP applied to the affected sequences; these situations do not yield non-conservative results.

Overall, the numerical impact of these event tree simplifications is judged to be very small. As a result, no specific action will be taken to address this finding.

RAI-30.01 Probabilistic Risk Assessment In your letter dated April 27, 2012 (ADAMS Accession No. ML12132A390) you responded to RAI-30, Probabilistic Risk Assessment, Uncertainty, discussing the results of sensitivity analyses using frequently asked question (FAQ) 08-0048 (ADAMS Accession No. ML092190457) fire ignition frequencies and the substantial risk reduction obtained from crediting a local operator action to restore the turbine driven auxiliary feedwater pump N train battery charger after it is tripped following loss of offsite power or after a safety injection signal. Given that including this action reduces the sensitivity study results to below the risk acceptance criteria in RG 1.174, discuss whether the RAI is or will be included in the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," fire protection program. If included, provide the additional risk of this RA.

If not included in the NFPA 805 fire protection program, provide justification for not including it.

Response to RAI 30.01 NFPA 805 Section 1.6.52 provides the following definition of Recovery Action (emphasis added):

"Activities to achieve the nuclear safety performance criteria that take place outside the main control room or outside the primary control station(s) for the equipment being operated, including the replacement or modification of components."

Regulatory Guide (RG) 1.205, Section C.2.4 states (emphasis added):

"Use of recovery actions, as defined in NFPA 805, Section 1.6.52, to demonstrate the availability of a success path for the nuclear safety performance criteria, does not meet the deterministic requirements in Section 4.2.3 of NFPA 805. Consequently, the licensee must address recovery actions, whether or not previously approved by the NRC, using the performance-based methods in Section 4.2.4, as required by NFPA 805, Section 4.2.3.1, and must evaluate the additional risk of their use according to NFPA 805, Section 4.2.4. Regulatory Position 2.2.4 provides guidance on calculating this additional risk of recovery actions."

to AEP-NRC-2013-1 Page 8 "NFPA 805, Section 4.2.3.1, identifies recovery actions for which the additional risk must be evaluated, as required by NFPA 805, Section 4.2.4. These success path recovery actions are operator actions that, if not successful, would lead to the fire-induced failure of the "one success path of required cables and equipment to achieve and maintain the nuclear safety performance criteria." Other operator actions that do not involve the success path may be credited in plant procedures or the fire PRA to overcome a combination of fire-induced and random failures may also be recovery actions, but licensees do not need to evaluate the additional risk of their use."

The above identified provisions from NFPA 805 and RG 1.205 indicate what a recovery action is and when additional risk from recovery actions is required to be evaluated.

The operator action to restore the N train battery charger is categorized by the last underlined statement above from RG 1.205. Restoration of the N train battery charger is a recovery action to overcome a combination of fire induced and random failures. It is not an NFPA 805 recovery action for the purposes of the CNP NFPA-805 License Amendment Request (LAR).

This action is currently included in CNP procedure OHP-4021-082-015, "Operation of the N Train Battery System," and included in the internal events PRA. The action is not included in the fire related procedures used in developing the CNP NFPA-805 LAR. This action is not an NFPA 805 recovery action because it is not needed to be compliant with NFPA 805 Nuclear Safety Performance Criteria.

For AFW pump related Variances from Deterministic Requirements (VFDR's), the specified NFPA 805 LAR recovery action to restore the AFW function is to cross-tie AFW from the unaffected unit.

Recovery actions for AFW cross-tie are provided in CNP fire procedures 1-OHP-4025-LS-2 and 2-OHP-4025-LS-2.

These are the NFPA 805 Recovery Actions credited in the Nuclear Safety Capability Assessment (NSCA) to ensure one success path free of fire damage.

Restoration of the N train battery charger is a desirable action in that it will provide a redundant train of AFW to mitigate random failures of the motor driven AFW trains. This action will only be effective for fire scenarios without fire damage to the turbine driven train, but with the following consequential events:

a) Safety Injection (SI) signal (spurious or real),

b) Loss of Offsite Power c) Failure of one train of AC power.

Restoration of the N train battery charger does not meet the definition of a recovery action in NFPA 805 Section 1.6.52. Other recovery actions are credited in NSCA to provide a success path.

Restoration of the battery charger is therefore not considered an NFPA 805 recovery action and does not need additional risk evaluation. Implementation item S-3.14 was added (by Reference 5) to Attachment S of the Transition Report to capture actions of this nature in appropriate operating procedures as part of the NFPA 805 implementation.

RAI-34.01 Probabilistic Risk Assessment In your letter dated August 9, 2012 (ADAMS Accession No. ML12242A246) and the non-public enclosure, you responded to RAI-34e, Probabilistic Risk Assessment, Peer Review F&Os, to AEP-NRC-2013-1 Page 9 discussing the results of sensitivity analyses that shows that using the "Special Weighting Factors" for transient fire frequency apportionment compared to using the nominal NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," (no fractional values) method will result in the delta core damage frequency (CDF) exceeding the RG 1.174 risk acceptance criteria. Provide the corresponding results for LERF and delta LERF.

Response to RAI 34.01 The August 9, 2012, response to RAI-34(e) provides the change (increase or decrease) in CDF due to three different sensitivity methods for transient frequency apportionment. This supplemental response provides the following for the three original sensitivity studies in our previous RAI-34(e) in addition to a sensitivity study using methods prescribed in draft FAQ 12-0064 (September 5, 2012):

a) Change in CDF b) Change in delta CDF c) Change in LERF d) Change in delta LERF The plant specific discussion of transient weighting factors provided in our previous response to RAI-34(e) will not be repeated in this response. The development of transient weighting factors from FAQ 12-0064 is summarized below:

a) A listing of 4000 work orders over the last seven years was retrieved and sorted to provide a relative ranking of maintenance activities in each of the plant fire zones.

b) A "50" weighting factor was assigned to three areas; one for turbine building, one for plant wide components and one for control/auxiliary building.

c) For these three areas, a "50" was assigned for hotwork and general maintenance d) For fire areas AA40, AA43, AA48, and AA52, a "0.3" was assigned for hotwork; a "1" for general maintenance and a "0.3" for storage. The assignment of these values is in accordance with I&M's commitment to make these areas transient combustible free and hotwork free zones. (See Attachment S, Items S-3.3 and S-3.4).

e) The control room cable vaults AA50, AA51 were assigned a "0" for hotwork, "0.3" for general maintenance, "0.3" for occupancy and "0" for storage. The assignment of these values is in accordance with I&M's commitment to make these areas transient combustible free and hotwork free zones, AND the natural accessibility restrictions of these areas.

Transient fire frequencies for the entire plant were recalculated using the 4-factor weighting scheme provided in FAQ-0064.

The results are shown in Table 34.01-1 through 34.01-8 Table 34.01-1 Unit 1 Change in Total CDF for Unit 1 Fire Areas Table 34.01-2 Unit 2 Change in Total CDF for Unit 2 Fire Areas to AEP-NRC-2013-1 Page 10 Table 34.01-3 Unit 1 Change in Total LERF for Unit 1 Fire Areas Table 34.01-4 Unit 2 Change in Total LERF for Unit 2 Fire Areas Table 34.01-5 Unit 1 Change in Delta CDF for Unit 1 Fire Areas Table 34.01-6 Unit 2 Change in Delta CDF for Unit 2 Fire Areas Table 34.01-7 Unit 1 Change in Delta LERF for Unit 1 Fire Areas Table 34.01-8 Unit 2 Change in Delta LERF for Unit 2 Fire Areas Tables 34.01-5 through 34.01-8 have a line which shows if the change in delta risk causes the plant to exceed risk acceptance criteria in RG 1.205. Meeting the risk acceptance criteria is shown for the base LAR results and when including the recovery actions for the TDAFP. This information is only shown for the delta risk tables because none of the total risk increases cause the plant to exceed RG 1.174 acceptance criteria.

RESULTS:

The first sensitivity study using the 1-1-1 minimum values as prescribed in NUREG/CR-6850 shows unacceptably high risk. However, that is to be expected, because NUREG/CR-6850 does not provide a numerical method to capture the effectiveness of transient combustible free zones or hotwork free zones.

The second and third sensitivity studies represent "unapproved methods" which were included in the first LAR response to provide a perspective for alternate methods. The results were included in this RAI, because they were requested. With the advent of the FAQ 12-0064 method, these sensitivities have no further use.

The FAQ 12-0064 sensitivity provides a reasonable basis to credit the effectiveness of procedures and restrictions to prevent transient fires. Using FAQ 12-0064 will cause a slight increase in CDF due to transient fires but does not prevent the plant from meeting RG 1.205 risk acceptance criteria.

to AEP-NRC-2013-1 Page 11 Table 34.01-1: Unit I Change in Total CDF for Unit I Fire Areas Transient Annortinninn Sensitivitv Studies Area Description BASE Values CDF change CDF change CDF change CDF change for Area when using when using when using when using 1-1-1 0-0-1

.05-.1-1 FAQ-0064 AA40 East 600V MCC

.05-.1-1 2.58E-07

-2.02E-08 NA 4.70E-08 ESF Room AA48 Aux Cable Vault

.05-.1-1 3.33E-06

-2.61 E-07 NA 6.06E-07 and Elect Switchgear Cable Enclosure AA50 Control Room

.05-.1-.2 1.29E-05 2.03E-06 2.80E-06

-2.69E-07 Cable Vault and Aux Shutdown Panel AA56 Containment 0-0-0 3.41 E-07 O.00E+00 2.75E-08 0.O0E+00 TOTAL 1.68E-05 1.75E-06 2.83E-06 3.84E-07 Table 34.01-2: Unit 2 Change in Total CDF for Unit 2 Fire Areas Transient Apportioning Sensitivity Studies Description CDF change CDF change CDF change CDF change BASE Values when using when using when using when using Area for Area 1-1-1 0-0-1

.05-.1-1 FAQ-0064 East 600V MCC

.05-.1-1 4.11E-08

-3.22E-09 NA 7.48E-09 AA43 ESF Room Control Room

.05-.1-.2 9.46E-06 1.48E-06 2.05E-06

-1.97E-07 Cable Vault and Aux Shutdown AA51 Panel Aux Cable Vault

.05-.1-1 9.87E-06

-7.73E-07 NA 1.80E-6 and Elect Switchgear Cable AA52 Enclosure AA58 Containment 0-0-0 2.06E-07 0.OOE+00 1.66E-08 0.OOE+00 TOTAL 1.96E-05 7.04E-07 2.07E-06 1.61 E-06 to AEP-NRC-2013-1 Page 12 Table 34.01-3: Unit I Change in Total LERF for Unit I Fire Areas Transient Apportioning Sensitivity Studies Area Description BASE Values LERF change LERF change LERF change LERF change for Area when using when using when using when using 1-1-1 0-0-1

.05-.1-1 FAQ-0064 AA40 East 600V MCC

.05-.1-1 6.77E-09

-5.30E-10 NA 1.23E-09 ESF Room AA48 Aux Cable Vault

.05-.1-1 1.94E-07

-1.52E-08 NA 3.54E-08 and Elect Switchgear Cable Enclosure AA50 Control Room

.05-.1-.2 1.71 E-6 2.67E-7 3.69E-7

-3.55E-08 Cable Vault and Aux Shutdown Panel AA56 Containment 0-0-0 1.17E-8 0.00E+00 9.43E-10 0.00E+00 TOTAL 1.92E-06 2.51E-07 3.70E-07 1.13E-09 Table 34.01-4: Unit 2 Change in Total LERF for Unit 2 Fire Areas Transient Apportioning Sensitivity Studies Description LERF change LERF change LERF change LERF change BASE Values when using when using when using when using Area for Area 1-1-1 0-0-1

.05-.1-1 FAQ-0064 East 600V MCC

.05-.1-1 3.17E-09

-2.48E-10 NA 5.77E-10 AA43 ESF Room Control Room

.05-.1-.2 1.27E-06 1.98E-07 2.74E-07

-2.64E-08 Cable Vault and Aux Shutdown AA51 Panel Aux Cable Vault

.05-.1-1 6.69E-07

-5.24E-08 NA 1.22E-07 and Elect Switchgear Cable AA52 Enclosure AA58 Containment 0-0-0 6.85E-09 0.OOE+00 2.69E-10 0.OOE+00 TOTAL 1.94E-06 1.46E-07 2.74E-07 9.62E-08 to AEP-NRC-2013-1 Page 13 Table 34.01-5: Unit I Change in Delta CDF for Unit I Fire Areas Transient AnDortionina Sensitivity Studies Area Description BASE Values Delta CDF Delta CDF Delta CDF Delta CDF for Area change when change when change when change when using using using using FAQ-1-1-1 0-0-1

.05-.1-1 0064 AA40 East 600V MCC

.05-.1-1 1.88E-07

-1.46E-08 NA 3.44E-08 ESF Room AA48 Aux Cable Vault

.05-.1-1 3.16E-06

-2.46E-07 NA 5.76E-07 and Elect Switchgear Cable Enclosure AA50 Control Room

.05-.1-.2 1.29E-05 2.03E-06 2.80E-06

-2.69E-07 Cable Vault and Aux Shutdown Panel TOTAL 1.62E-05 1.77E-06 2.80E-06 3.41 E-07 OVER RG 1.205 LIMIT IN BASE CASE LAR YES YES YES NO OVER RG 1.205 LIMIT W/CREDIT FOR TD-AFW YES NO YES NO Table 34.01-6: Unit 2 Change in Delta CDF for Unit 2 Fire Areas Transient Apportioning Sensitivity Studies Description Delta CDF Delta CDF Delta CDF Delta CDF change when change when change when change when BASE Values using using using using FAQ-Area for Area 1-1-1 0-0-1

.05-.1-1 0064 East 600V MCC

.05-.1-1 3.28E-08

-2.55E-09 NA 5.98E-09 AA43 ESF Room Control Room

.05-.1-.2 9.46E-06 1.48E-06 2.05E-06

-1.97E-07 Cable Vault and Aux Shutdown AA51 Panel Aux Cable Vault

.05-.1-1 6.67E-06

-5.19E-07 NA 1.22E-06 and Elect Switchgear Cable AA52 Enclosure TOTAL 1.62E-05 9.58E-07 2.05E-06 1.03E-06 OVER RG 1.205 LIMIT IN BASE CASE LAR YES NO YES NO OVER RG 1.205 LIMIT W/CREDIT FOR TD-YES NO NO NO AFW to AEP-NRC-2013-1 Page 14 Table 34.01-7: Unit I Change in Delta LERF for Unit I Fire Areas Transient Apportioning Sensitivity Studies Area Description BASE Values Delta LERF Delta LERF Delta LERF Delta LERF for Area change when change when change when change using using using when using 1-1-1 0-0-1

.05-.1-1 FAQ-0064 AA40 East 600V MCC

.05-.1-1 1.71E-09

-1.33E-10 NA 3.13E-10 ESF Room AA48 Aux Cable Vault

.05-.1-1 1.84E-07

-1.48E-08 NA 3.35E-08 and Elect Switchgear Cable Enclosure AA50 Control Room

.05-.1-.2 1.71 E-06 2.67E-07 3.69E-07

-3.55E-08 Cable Vault and Aux Shutdown Panel TOTAL 1.90E-06 2.52E-07 3.69E-07

-1.69E-09 OVER RG 1.205 LIMIT IN BASE CASE LAR YES NO YES NO OVER RG 1.205 LIMIT W/CREDIT FOR TD-YES NO NO NO AFW Table 34.01-8: Unit 2 Change in Delta LERF for Unit 2 Fire Areas Transient Apportioning Sensitivity Studies Description Delta LERF Delta LERF Delta LERF Delta LERF change when change when change when change when BASE Values using using using using FAQ-Area for Area 1-1-1 0-0-1

.05-.1-1 0064 East 600V MCC

.05-.1-1 2.24E-09

-1.74E-10 NA 4.09E-10 AA43 ESF Room Control Room

.05-.1-.2 1.27E-06 1.98E-07 2.74E-07

-2.64E-08 Cable Vault and Aux Shutdown AA51 Panel Aux Cable Vault

.05-.1-1 4.58E-07

-3.56E-08 NA 8.35E-08 and Elect Switchgear Cable AA52 Enclosure TOTAL 1.73E-06 1.62E-07 2.74E-07 5.75E-08 OVER RG 1.205 LIMIT IN BASE CASE LAR YES NO NO NO OVER RG 1.205 LIMIT W/CREDIT FOR TD-YES NO NO NO AFW