NRC-2014-0043, Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805

Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805

ML14210A645
Person / Time
Site:	Point Beach
Issue date:	07/29/2014
From:	Mccartney E Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-2014-0043
Download: ML14210A645 (93)

Text

NEXTera**

ENERGY~

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July 29, 2014 NRC 2014-0043 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805

References:

(1) NextEra Energy Point Beach, LLC, letter to NRC, dated June 26, 2013, "License Amendment Request 271, Transition to 10 CFR 50.48(c)-

NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition" (ML131820453)

(2) NRC e-mail to NextEra Energy Point Beach, LLC, dated September 9, 2013, "Request for Supplemental Information Regarding the Acceptability of the Proposed Amendment Request" (ML13256A197)

(3) NextEra Energy Point Beach, LLC, letter to NRC, dated September 16, 2013, "License Amendment Request 271 Supplement 1 Transition to 10 CFR 50.48(c)- NFPA 805" (ML13259A273)

(4) NRC letter to NextEra Energy Point Beach, LLC, dated September 25, 2013, "Point Beach, Units 1 and 2- Acceptance Review of Licensing Action re:

License Amendment Request to Transition to NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (ML13267A037)

(5) NRC e-mail to NextEra Energy Point Beach, LLC, dated July 8, 2014, "Point Beach Nuclear Plant, Units 1and 2 - Final (Revised) Requests for Additional Information re: License Amendment Request Associated with NFPA 805 (TAC Nos. MF2372 and MF2373)" (ML14189A365)

Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC, (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Reference 1 and supplemented via Reference 3). The NRC accepted the license amendment request for review in response to Reference 2 as documented in Reference 4.

The NRC Staff has determined that additional information (Reference 5) is required to complete its evaluation. Enclosure 1 provides the NextEra response to the NRC Staff's request for additional information for the required 60 Day Response.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, Wl54241

This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on July 29, 2014.

Very truly yours, NextEra Energy Point Beach, LLC v~/11C0~y Eric McCartney , - ~0 Site Vice President Attachments A - G cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 RESPONSE (60 DAY) TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST ASSOCIATED WITH NFPA 805 Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC, (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Reference 1 and supplemented via Reference 3). The NRC accepted the license amendment request for review in response to Reference 2 as documented in Reference 4.

The NRC Staff has determined that additional information (Reference 5) is required to complete its evaluation. This Enclosure 1 provides the NextEra response to the NRC Staff's request for additional information for the 60 Day Response.

Probabilistic Risk Assessment (PRAl RAJ 05 - Transient Heat Release Rate CHRR)

NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1. 205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Rev. 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA 805. Methods that have not been determined to be acceptable by the NRC Staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC Staff to complete its review of the proposed method.

LAR Attachment V,Section V.2 indicates that the transient fire for the cable spreading room (CSR) is characterized by a 142 kilowatt (kW) fire and the vital switchgear room is characterized by a 69 kW fire. Both reduced HRRs are stated to be justified because they are credited in designated enhanced transient combustible compartments. The staff determined that the justification provided in LAR Section V. 2 is insufficient for the staff to complete its review.

Discuss the key factors used to justify the credit for reduced transient fire HRRs below 317 kW including:

a) Identification of any other fire compartments for where reduced HRR transient fires are credited; b) For each location where a reduced HRR is credited, a description of the administrative controls that justify the reduced HRR including how location-specific attributes and considerations are addressed. Provide a discussion of required maintenance for ignition sources in each location, and types/quantities of combustibles needed to perform that maintenance. Also discuss the personnel traffic that would be expected through each location; c) The results of a review of records related to violations of the transient combustible and hot work controls; and, d) A discussion of the impact on the analysis.

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NextEra Response a) There are no other fire compartments that utilized a transient combustible heat release rate (HRR) less than the 981h percentile HRR (i.e., 317 kW), as specified in NUREG/CR-6850, Table G-1.

b) Cable Spreading Room- Fire Compartment 318 The 751h percentile transient combustible heat release rate (142 kW) was implemented in the Cable Spreading Room (Fire Compartment 318). The Cable Spreading Room contains 480V load centers, 125V distribution panels, 120V instrument panel transformers (dry-type), 120V inverters, and station service transformers (containing PCBs). The combustibles in this fire compartment are limited to cable insulation; however, all cable trays in the Cable Spreading Room are fully enclosed with heavy gauge steel and %" Kaowool insulation is provided on top of the cables below the top cover. Although some of the transformers contain PCBs, large combustible liquid fires are not expected in these areas. PCBs are combustible; however, they do not sustain a flame in the absence of a pilot source in the way lube oil or diesel fuel would.

The 751h percentile transient heat release rate (142 kW) was selected for this fire zone based on large combustible liquid fires not being expected in the zone, since pumps are not located in the area and activities in the fire zone do not include maintenance of oil containing equipment. A review of the routine preventative maintenance activity work orders indicate that the typical activities are breaker maintenance, thermographic inspection of inverters, functional testing of relays, calibration and bench test of relays, infrared buswork inspection, ammeter calibration, transformer thermography, fire alarm control panel battery replacement, oil sampling of transformers, and ductwork inspection. Maintenance procedures referenced in the work orders provide the measurement and test equipment, and special tools required for completing the activities. The equipment and tools typically consist of multimeters, oscilloscopes, panel test meter, grounding strap, low voltage gloves, penetration caps, relay test base, extension cord, torque wrenches, torque screwdriver, megger, ohmmeter, spring scale, feeler and pin gauges, electrical breaker maintenance tool box, 9 volt battery, clean cloth, isopropyl alcohol, graphite grease, and fire blankets.

The types and quantities of combustibles needed to perform these activities consist of limited amounts of plastic parts on the equipment and a plastic cart is typically used for transporting equipment in and out of the area. Although, a plastic cart and other combustibles are currently utilized as part of these maintenance activities, their use requires constant attendance by the maintenance personnel or a continuous fire watch. Additionally, all combustibles are required to be removed from the area upon job completion.

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Per Procedure NP 1.9.9, "Transient Combustible Control," the Cable Spreading Room is recognized as a fire area that requires a heightened level of awareness for controlling transient combustibles to minimize potential fire risk. These controls include combustible materials being stored in closed solid metal containers, evaluation performed by qualified Fire Protection staff or a continuous fire watch required for excess combustibles, and all combustibles to be removed from the fire zone upon job completion. Monitoring for proper transient combustible controls is conducted by performing safe shutdown area tours, performing housekeeping tours, and providing compensatory measure fire watches. Procedure NP 1.9.9 has been revised in accordance with Implementation Item IMP-144 in Table S-3 of the LAR (Reference 1), which establishes enhanced transient and combustible controlled zones in high risk Fire Compartment 318 (Cable Spreading Room). These controls include verification that appropriate fire detection and suppression are in service and that no hot work activity is allowed in the fire zone without prior approval.

The Cable Spreading Room has restricted security access requiring card reader access through the locked doors. The Cable Spreading Room is accessible via two personnel doors from the non-vital switchgear room (northeast and southeast corner).

There is also a spiral staircase providing access between the Cable Spreading Room and the Control Room above. Based on the configuration and restricted access of this zone, personnel traffic is expected to be low and limited to routine maintenance activities, as this area is not a common pathway to an adjacent area.

Vital Switchgear Room - Fire Compartment 305 A transient heat release rate of 69 kW was implemented in the Vital Switchgear Room (Fire Compartment 305). The Vital Switchgear Room contains 4160V switchgear and 125V DC distribution panels. The combustibles in this fire compartment are limited to cable insulation and minor floor-based combustible loading.

A 69 kW transient heat release rate was selected for the Vital Switchgear Room based on large combustible liquid fires not being expected in the zone, since pumps are not located in the area and activities in the fire zone do not include maintenance of oil containing equipment. A review of the routine preventative maintenance activity work orders indicates that the typical activities are switchgear infrared survey, breaker maintenance, calibration of relays, buswork inspection, charger inspection and maintenance, station battery charger thermography, battery charger ammeter inspection, mechanism-operated cell switch fulcrum plate mounting inspection, flow switch cleaning, fan bearing lubrication and inspection, and seismic event indicator test. The types and quantities of combustibles needed to perform these activities consist of limited amounts of plastic parts on the equipment and a plastic cart is typically used for transporting equipment in and out of the area. Although, a plastic cart and other combustibles are currently utilized as part of these maintenance activities, their presence requires constant attendance by the maintenance personnel or a continuous fire watch. Additionally, all combustibles are required to be removed from the area upon job completion.

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Per Procedure NP 1.9.9, "Transient Combustible Control," the Vital Switchgear Room is recognized as a fire area that requires a heightened level of awareness for controlling transient combustibles to minimize potential fire risk. These controls include combustible materials being stored in closed solid metal containers, evaluation performed by qualified Fire Protection staff or continuous fire watch required for excess combustibles, and all combustibles to be removed from the fire zone upon job completion. Monitoring for proper transient combustible controls is conducted by performing safe shutdown area tours, performing housekeeping tours, and providing compensatory measure fire watches. Procedure NP 1.9.9 has been revised in accordance with Implementation Item IMP-144 in Table S-3 of the LAR (Reference 1), which establishes enhanced transient and combustible controlled zones in high risk Fire Compartment 305 (Vital Switchgear Room). These controls include verification that appropriate fire detection and suppression are in service and that no hot work activity is allowed in the fire zone without prior approval.

This fire compartment has restricted security access requiring card reader access through the locked doors. The Vital Switchgear Room is accessible via two personnel doors, one in the south wall from the Unit 1 turbine hall and one in the northwest wall from the remote shutdown panel area. Based on the configuration and restricted access of this zone, personnel traffic is expected to be low and limited to routine maintenance activities, as this area is not a common pathway to an adjacent area.

c) Records dated between May 2010 and May 2014 identifying violations of hot work and transient combustible controls were reviewed. During this period, for Fire Compartment 305 and Fire Compartment 318, there were no Condition Reports (CRs) written for hot work control issues and only three CRs written for minor transient combustible control issues for the Cable Spreading Room, which are further discussed below. A description of each transient combustible control issue is shown in the table below. None of the identified CRs resulted in a fire and the combustibles identified in the one CR would not be expected to exceed the reduced heat release rate modeled in the fire PRA. Additionally, a continuous fire watch was implemented as a compensatory measure for the use of these combustibles.

Table 1: CRs Created for Transient Combustible Control Issues Fire CR Date Description of Transient Combustible Control Issues Compartment Electrical Maintenance requested routing of a temporary power cable through the Cable Spreading Room. The FPE directed the cable to be routed on the floor and enclosed for its entire length with flexible conduit or wrapped with fire blanket. However, 10/15/13 318 additional combustible items were added without further compensatory measures. These items included a wood table, plastic carts, paperwork, plastic bags, and extension cords. Once identified, a continuous fire watch was put in place.

Materials identified on a Transient Combustible Control Form 12/01/11 318 were found in several areas, including the Cable Spreading Room, but no form was physically posted at the materials.

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Table 1 = CRs Created for Transient Combustible Control Issues Fire CR Date Description of Transient Combustible Control Issues Compartment Temporary storage permits for transient combustibles in several areas, including the Cable Spreading Room, were identified to be 3/25/11 318 past their expected completion date. Upon review of the Cable Spreading Room, all transient combustibles were removed and the temporary storage permits were closed.

d) The review of the CRs performed in response to RAI PRA 05.c identified that none of the transient violations resulted in a fire and the combustibles identified in the CRs would not result in a transient fire with a heat release rate exceeding the reduced heat release rates modeled in the fire PRA. The review of the required preventative maintenance for ignition sources in the Vital Switchgear Room and the Cable Spreading Room performed in response to RAI PRA 05.b, identified that the minimal quantities of combustibles needed to perform the activities would not typically exceed the reduced heat release rates modeled in the fire PRA. In the event that combustibles are required to be brought into these areas, their presence will require a continuous fire watch and all combustibles are required to be removed from the area upon job completion.

Therefore, based on the enhanced administrative controls that will be put in place for the transition to NFPA 805, the limited personnel traffic expected in these areas, and minimal combustibles required for maintenance activities in these areas, the reduced heat release rates (69 kW and 142 kW) were determined to be appropriate to represent the expected combustibles in these areas.

PRA RAI 06 - Use of Unacceptable Methods NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method. LAR Attachment V,Section V. 2 states that the "PBNP fire PRA did not use unreviewed analysis methods."

Indicate if any methods not yet accepted by NRC were used. If so, provide a summary of those methods in sufficient detail to enable the NRC staff to complete its review. Also, determine the impact on fire CDF, LERF, llCDF, and llLERF for those methods.

NextEra Response At the time of the LAR submittal, the statement provided in LAR Attachment V,Section V.2, "PBNP fire PRA did not use unreviewed analysis methods," was accurate. However, since the LAR submittal, the NRC has provided updated guidance for several methods acceptable to the NRC Staff.

The details and impacts of the updated method guidance are being addressed directly through other RAI responses, specifically PRA RAis 1.a, 1.e and 4. These RAI responses will be provided in the 120 day response.

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PRA RAJ 07~ Transient Fire Placement at Pinch Points NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG!CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff require additional justification to allow the NRC staff to complete its review of the proposed method.

The NRC staff could not determine how pinch points" were modeled for transient fires.

Transient fires should at a minimum be placed in locations within the plant physical analysis units (PAUs) where conditional core damage probabilities (CCDPs) are highest for that PAU, (i.e., at pinch points'1. Pinch points include locations of redundant trains or the vicinity of other potentially risk-relevant equipment. Hot work should be assumed to occur in locations where hot work is a possibility, even if improbable, keeping in mind the same philosophy.

a) Clarify how pinch points" were identified and modeled for transient fires.

b) Describe how transient and hot work fires are distributed within the PAUs. In particular, identify the criteria used to determine where such ignition sources are placed within the PAUs.

NextEra Response a) During the fire modeling analysis, risk-significant transient fire areas were identified based on any of the following criteria:

• Locations where a transient fire could fail PRA equipment.

• Locations where a transient fire could fail PRA cable trays or conduit.

• Locations with a large number of secondary combustibles that could lead to damaging smoke plumes, radiant damage, or hot gas layer damage to PRA equipment, cables, or conduit.

Using any combination of the above criteria, the boundaries of each transient zone are chosen such that the associated fire growth and resulting damage to PRA targets (i.e., cables and equipment) can be bounded by a representative fire scenario. These locations were identified such that the modeled fire conditions represent the worst case damaging conditions (i.e., largest target sets) for a given compartment.

Therefore, the size of each transient zone will be larger than the postulated zone of influence for a single transient fire. Using the approach that transient and hot work fires are postulated anywhere a transient fire is reasonably expected to occur (all accessible floor areas except where precluded by design and/or operation), specific "pinch points" need not be identified as they will be conservatively bound by these representative scenarios.

b) Transient and hot work fires are distributed within the plant physical analysis units (PAUs) in accordance with the process described in EPM Procedure EPM-DP-FP-001, "Detailed Fire Modeling", Revision 3, Section 7.6.2. To summarize:

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In general, transient and hot work fires are postulated anywhere a transient fire is reasonably expected to occur. This is typically all accessible floor areas except where precluded by design and/or operation (e.g., plant equipment). The accessible floor area of each PAU is then subdivided into one or more transient zones. The boundaries of each transient zone are chosen such that the associated fire growth and resulting damage to PRA targets (i.e., cables and equipment) can be bounded by a representative fire scenario.

In order to keep the number of locations (and therefore the number of transient scenarios) requiring separate analysis to a minimum, PAUs have transients placed in locations where risk significant targets, or target combinations are located. Transient fire locations within the immediate damage range of these targets are then defined, and the transient ignition frequency apportioned to these locations. The remainder of the floor space of the PAU is subdivided only where it is necessary to distinguish between different fire growth potential (e.g., locations where secondary combustibles are at a low enough elevation to be ignited by the transient fire).

PRA RAI10- Main Control Room Abandonment Modeling NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

LAR Attachment W, Section W.2.1 indicates that the FPRA credits MCR abandonment for both habitability and non-habitability scenarios. The licensee's analysis appears to indicate that MCR abandonment cases are represented by a single cutset (with a CDF of 6. 73E06/year for Unit 1 and 6. 79E-06/year for Unit 2). The cutset frequencies presented in the analysis are the same as the scenario frequencies presented the MCR abandonment scenario in LAR Attachment W, Tables W-2 and W-3, indicating that the abandonment scenario consists of just a single cutset. The licensee's analysis explains that single CCDP values (i.e., 0.65 for Unit 1 and

0. 66 for Unit 2) were determined for MCR abandonment scenarios by summing the six HEPs (per unit) associated with actions required for alternate shutdown. The analysis further explains that the contributions from random equipment failures were assumed to be dominated by the HEPs.

Also, given the discussion in the licensee's analysis, it is not clear why the cited CCDP values are not presented as contributors to the MCR abandonment cutset presented in the licensee's analysis.

Also, it is not clear whether the six HFEs (per unit) cited in the MCR analysis are adequate to represent the actions required to estimate the CCDP for the range of equipment failures for abandonment. LAR Attachment G, Table G-1 seems to indicate there are a number of other actions required for alternate shutdown (e.g., recovery actions associated with components 1 L/-426, 1 PI-420C, 1 PI-438A, 1T/-451 B, 1T/-451 C, 2LI-426, 2L/-470A, 2N/-00040, 2PI-420C, 2T/-451 B, and 2TI-451 C).

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LAR Attachment W, Section W.2 a/so presents nine MCR abandonment failures used to help calculate the CCDP of the compliant plant for "non-habitability" cases in Fire Areas A24, A30, and A31. It is not clear whether these failures are the same or related to the abandonment HFEs discussed in the licensee's analysis. Abandonment actions appear not be credited for Fire Area 24 (4 kV switchgear room) or for Fire Area A30 (CSR) because there is no abandonment scenario presented for these fire areas in the corresponding analysis. Therefore, it appears, though it is not clear, that abandonment actions were credited in the compliant case but not the post transition case.

Accordingly, it is not completely clear from the documentation how MCR abandonment was treated in the FPRA, how the scenario frequency for MCR abandonment was determined, or how potential fire-induced failures resulting from fires leading to MCR abandonment were addressed. It appears that single CCDPICLERP values were used to represent a range of potential MCR abandonment scenarios. In light of the observations, provide the following:

a) Describe how MCR abandonment was modeled for loss of habitability. Include identification of the actions required to execute safe alternate shutdown and how they are modeled in the FPRA, including actions that must be performed before leaving the MCR. Also, include an explanation of how the CCDPs and CLERPs are estimated for fires that lead to MCR abandonment.

b) Explain how the CCDPs and CLERPs estimated for fires that lead to abandonment due to loss of habitability address various possible fire-induced failures. Specifically include in this explanation, discussion of how the following scenarios are addressed:

f. Scenarios where fire fails only a few functions aside from forcing MCR abandonment and successful alternate shutdown is straightforward; H. Scenarios where fire could cause some recoverable functional failures or spurious operations that complicate the shutdown, but successful alternate shutdown is likely; and, iii. Scenarios where the fire-induced failures cause great difficulty for shutdown by failing multiple functions and/or complex spurious operations that make successful shutdown unlikely.

c) Explain how the abandonment scenario frequency due to loss of habitability was determined.

Include explanation of how the fire ignition frequencies contributing to this scenario and non-suppression probabilities were addressed.

d) Explain whether MCR abandonment is being credited for non-habitability cases (i.e., loss of control in the MCR). Based on LAR Attachment IN, Section W.2.1, it appears to be credited only in the compliant case.

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Indicate if abandonment due to loss of control is credited for the post transition case. If abandonment due to loss of control is being credited in the post transition case then describe the scenarios to which this applies. Address parts a) and c), above, for abandonment due to loss of control. Also describe the cues used by operators to abandon the MCR and how the timing of these cues is determined and modeled. Include an explanation of how the combinations of different functions that can be lost due to different fire-induced failures are addressed in the FPRA modeling. Note that absent an acceptable analysis for abandonment due to loss of control, the staff will not accept the credit for this particular type of recovery.

Also, referring to the questions asked for the post transition case, describe the scenarios and analysis which apply to the compliant case.

NextEra Response a) Main Control Room (MCR) abandonment for loss of habitability was modeled based on the failure of operators to successfully execute a minimum set of actions that were identified to accomplish safe shutdown outside of the MCR. The actions required are identified in "Alternate Shutdown Human Reliability Analysis" (P2091-000-001, Rev. 0), which is based on the main procedural steps and procedural steps in attachments A through D of AOP-10A, "Safe Shutdown- Local Control." The human error probabilities (HEPs) for these events are also developed in the "Alternate Shutdown Human Reliability Analysis" calculation. A summary of the required events are as follows:

• Fire recognized in the control room

• Steps 1-4 of EOP-0, "REACTOR TRIP OR SAFETY INJECTION," are performed before Shift Manager enters AOP-10A as a conservative assumption for timing

• Aligning equipment and placing hand switches in pull-to-lock prior to abandonment per AOP-10A, "SAFE SHUTDOWN- LOCAL CONTROL." These steps provide defense-in-depth because alignments and isolations are verified by the ex-control room actions

• Perform ex-CR actions as per AOP-10A to establish safe shutdown As per Section 10 of the "NFPA 805 Fire ... [PRA] Main Control Room Analysis" (P2091-2700-01, Rev. 1), the HEPs are summated to be used as conditional core damage probabilities (CCDP) values, which are 5.65E-01 and 5.66E-01 for Unit 1 and Unit 2, respectively. Conditional large early release probabilities (CLERPs) are assumed to be one level of magnitude lower for both units. Hardware failures were not included in these calculated values, as the HEPs were conservatively developed and assumed to provide a bounding estimate for the CCDP and CLERP values for MCR abandonment.

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b) The full main control board (MCB) fire ignition frequency is applied to both abandonment and non-abandonment scenarios for MCB fire (to be addressed in response to RAI PRA 3 and RAI PRA 9 in the 120 Day RAI Response). The non-abandonment conditional core damage probabilities (CCDPs) and conditional large early release probabilities (CLERPs) are determined by evaluating each ignition source in the main control room (MCR) and their contribution to core damage frequency (CDF) by failing the source and associated targets as per the methodologies provided in Section 2 of the "NFPA 805 Fire Probabilistic Risk Assessment Quantification Notebook" (P2091-2900-02, Rev. 1). The target sets are determined in Section 7 of the "NFPA 805 Fire PRA Main Control Room Analysis (P2091-2700-01, Rev. 1). No split fractions are used to credit abandonment. In some cases, this is conservative as there is some possibility that the operators decide to abandon in time to mitigate core damage.

Abandonment was only credited for loss of habitability. As per Section 11.5.2 and Appendix J of the "NFPA 805 Fire PRA Main Control Room Analysis" a severity factor (SF) and non-suppression probability (NSP) are determined for a diverse set of heat release rates (HRR) and configurations. After convolution, (SF*NSP), the source ignition frequency and abandonment CCDP [as described in the response to RAI PRA 1O(a)] are multiplied together to determine the abandonment CDF for each scenario. The human action steps in the attachments of AOP-1 OA to recover plant equipment caused by fire failures in the control room are considered required steps for success.

No other reduction factors are considered in the abandonment or non-abandonment cases. This treatment captures all failure modes in the control room for abandonment and non-abandonment conservatively. Therefore, each of the scenarios listed in this RAI (i, ii, iii) are bounded by this application, rather than addressed individually.

c) The frequency of abandonment is derived from the non-suppression probability (NSP) at the time the abandonment threshold is reached. As described in the response to RAI PRA 10 (b),

an abandonment time was determined for each ignition source based on fire growth and room configuration. The abandonment thresholds are based on environmental factors reaching habitability limits (Section 11 of "NFPA 805 Fire PRA Main Control Room Analysis," P2091-2700-01 Rev. 1). The criteria are defined as the following:

• The heat flux at 6 feet above the floor exceeds 1 kW/m 2

• The smoke layer descends below 6 feet from the floor, and optical density of the smoke is greater than 3.0 m-1

• The hot gas layer at 6 feet above the floor exceeds 95°C (200°F)

• A fire inside the main control board damaging internal targets 2.13 m (7') apart The factors influencing the timing are heat release rates (HRR) for initiators from Appendix E in NUREG/CR-6850, availability of smoke exhaust and control room configuration. The timing factors are used for calculation on the NSP. Severity factors are also obtained from NUREG/CR-6850. A product of the fire ignition frequencies, severity factors (SF), NSP and abandonment conditional core damage probabilities (CCDP) developed as per the response to RAI PRA 10 (a) are used to develop the core damage frequency (CDF) for each scenario.

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d) For the post-transition model, main control room (MCR) abandonment is not being credited for non-habitability cases. Therefore, the post-transition model does not credit loss of control.

In the compliant case, MCR abandonment is credited in all scenarios (see Attachment W of the LAR - Reference 1). In all cases, it was assumed that the operator actions were always successful, which is consistent with pre-transition deterministic assumptions for alternate shutdown. However, equipment failures were considered. G-05 (gas turbine) and the turbine driven auxiliary feed water pump are assumed to be used to shut the plant down. Based on the summation of G-05 and turbine driven auxiliary feed water pump related basic event failure probabilities, a conditional core damage probability (CCDP) of 0.19 and conditional large early release probability (CLERP) of 0.019 (one level of magnitude lower from CCDP) are developed.

In regards to RAI PRA 10 a, band c, the methodologies of modeling for habitability and calculating abandonment scenario frequency do not change for the compliant case. CCDP and CLERP values are the only major differences from the post-transition model, which is previously discussed in this RAI response.

PRA RAI 11 - Command and Control of MCR Abandonment Actions NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

LAR Attachment G states that PBNP does not have any locations considered to be Primary Control Stations (PCSs). It is not clear where or how centralized command, control, and communication are performed for MCR abandonment actions. Explain how the feasibility of command and control, and communication and coordination of actions related to MCR abandonment was determined and is factored into the FPRA.

NextEra Response Point Beach does not have a Primary Control Station (PCS) for main control room (MCR) abandonment as defined by Revision 1 of RG 1.205. In the event of a severe MCR fire, the "Safe Shutdown- Local Control" procedure (AOP-1 OA) will be executed by Operations. The Duty Operating Supervisor (DOS) provides command and control of AOP-10A, communicating to operators using Gai-tronics and/or portable radio as they complete critical steps in the attachments of AOP-10A. The DOS also acts as the coordinator during execution of local alignment actions in Attachment A of AOP-10A as he/she will receive regular feedback from operators performing other attachments in the procedure.

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The feasibility of MCR abandonment actions is documented in Report R2168-1 012-01 Revision 0, "Manual Actions Feasibility Evaluation"; all actions associated with MCR abandonment are considered "Recovery Actions" (RAs) in accordance with NFPA 805 Section 1.6.52, FAQ 07-0030, and RG 1.205 Rev. 1. The feasibility assessment is based on a documentation review of the existing plant evaluation (FPTE 012) and plant walkdowns. The results support that all of the credited NFPA 805 Recovery Actions are feasible, and operations training/drills demonstrates the adequacy of command and control aspects. Operators are subject to Job Performance Measures (JPMs) for local control of the Chemical and Volume Control System (CVCS) and the turbine-driven auxiliary feed-water pump (TDAFWP). Additionally, AOP-10A is reviewed and exercised periodically during classroom rotations.

The RAs are currently credited in existing post-fire response procedures (AOP-1 OA and FOP 1.2, "Potential Fire Affected Safe Shutdown Components") and Operating Instructions (e.g., 01 110, "Gas Turbine Operation"), under the existing 10 CFR 50 Appendix R Fire Protection Licensing Basis, and have been validated using the "EOP/AOP VerificationNalidation Process" (OM 4.3.2). Note that FOP 1.2 directs the operator(s) into AOP-10A and other procedures for non-MCR abandonment scenarios. Additional actions for MCR abandonment (not part of the FPTE 012 timeline for G-05, gas turbine-driven generator usage, but part of AOP-1 OAf FOP 1.2 or new PRA calculations) were individually walked down by PBNP operations staff as per step 4 of the "Manual Actions Feasibility Evaluation."

Communications (Gai-tronics and portable radios) are evaluated in Attachment 2 of R2168-1012-01 Revision 0 to demonstrate that, in the case of MCR abandonment, a mode of communication is available to Operations and the fire brigade. Point-to-point communications (such as AFW alignment to CVCS alignment communication) have also been evaluated in FPTE 007, 'Technical Evaluation of PBNP Point-to-Point Portable Radio Communications for an Appendix R Fire," for feasibility. In addition, the procedure prioritizes restoring power to non-safeguards 480V Motor Control Centers (MCCs ), B-33 and B-43, for radio repeaters to improve communication in non-tested areas.

As described in Section 4.6 of NUREG-1921, Human Reliability Analysis (HRA), the FPRA accounts for crew communications, crew dynamics, complexity and other variables in the form of performance shaping factors (PSFs) to determine the probability of success of a given action or group of actions. The FPRA considers PSFs and variables related to feasibility of command and control, and communication and coordination of actions of MCR abandonment in the "Alternate Shutdown Human Reliability Analysis" (P2091-000-001, Revision 0). A high stress PSF is applied to each developed human recovery action developed to represent MCR abandonment, which accounts for the complexfty of the organizational management, remote communications and unusual plant conditions.

PRA RAI 12- Use of Assumed Cable Routing NFPA 805 Section 2.4.3.3 states that the PRA approach methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

a} The licensee's analysis states that "Systems without cable tracing are failed unless further analysis was petformed to assure systems are not compromised by the transient or fire (credit by exception}."

i. Explain how "credit by exception" was petformed for untraced cables.

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ii. Also, identify the functions and systems (including loss of instrumentation and control functions) which are assumed to be failed because of lack of information about cable routing. Describe how these assumed failures impact the calculated risk values in the post-transition and the compliant PRA models. Include explanation of whether they impact the post-transition and compliant models differently, and the implications of those differences on the change in risk estimates.

b) The NRC staff is concerned with regards to conservative PRA assumptions in the case where the change in plant risk due to a modification can be significantly overestimated.

The NRC staff notes, for example, that failure to trace cable and credit the unaffected train in the compliant plant can overestimate the risk of the compliant plant and consequently, the risk benefit from a plant modification, such as the addition of a train for injection/decay heat removal redundant or diverse to the assumed failed train, can be significantly overestimated.

i. Indicate whether additional basic events credited in the PRA could change dominant conservative scenarios in the complaint plant to negligible contributors in the post transition plant.

ii. If cases are identified where credit for a plant modification is overestimated from the stated concern, provide a discussion of the realism associated with the risk reduction for the particular plant modification.

NextEra Response a.i) "Credit by exception" was not used in the Point Beach Fire PRA.

a.ii) There are some mitigating events modeled in the internal events PRA, which are failed in the fire PRA. They generally are grouped into the following categories:

• Component events that do not support fire mitigation scenarios (e.g., events only applicable to steam generator tube rupture, ATWS, Steam and Feed Water line breaks inside containment, etc.)

• Events that have limited importance to fire scenarios that are qualitatively addressed (e.g., back feeding power and alternate offsite power). Qualitative impacts on results are presented in the Table in response to Part b) i. of this response.

• Events that are not credited for fire due to high failure likelihood (i.e. Main Feed Water (MFW)). Qualitative impacts on results are presented in the Table in response to Part b.i) of this response.

b.i) The failed components were reviewed and grouped into major mitigating functions. The following table provides dispositions of the impact of identified failed functions in the fire Probabilistic Risk Assessment (FPRA).

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Group Disposition Function Instrument Air is failed due to lack of high temperature piping justification. To Instrument avoid a conservative delta risk calculation Instrument Air is not failed for the and Service compliant model. Fire impacts may still fail air for the compliant case, which Air will be further discussed in the response to PRA RAI-01.d in the 120 Day RAI Response.

These events support ATWS mitigation, which are screened scenarios in the ATWS FPRA.

Continued operation of main feed water (MFW) requires availability of a large number of circuits that are located in a large percentage of the plant. The main feed water system would only be expected to be beneficial in areas where auxiliary feed water (AFW) mitigation is significantly damaged. The Main Feed primary location where additional feed water system support would be Water beneficial would be the AFW pump rooms, vital switchgear room and cable spreading room; these areas are expected to have main feed water circuits.

Therefore any impact of modeling MFW in the fire PRA is expected to be limited.

Back feed is expected to have limited risk reduction based on it being a lengthy evolution that requires human action outside the main control room Back feed and cable routing is expected to share locations with normal post trip offsite power circuits.

Alternate Offsite Power Alternate offsite power feed to the non-safety 4160V buses is expected to Feeds for have limited risk reduction due to shared cable routing location with currently Non-Safety credited sources power.

4160V Buses Steam Generator These events support SGTR mitigation, which are screened scenarios in the Tube FPRA.

Rupture (SGTR)

Steam and Feed Line These events support Feed and Steam line breaks mitigation, which are break inside screened scenarios in the FPRA.

containment b.ii} Except for Instrument Air, which is further described in the response to PRA RAI-01.d (PRA RAI-01.d is to be submitted with the 120 Day Response), no significant over-estimation of risk is expected due to assumed failures in the FPRA based on the response to part b.i of this RAI response.

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PRA RAI13 *Fire PRA Credit for Westinghouse RCP Seals NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4. 1 of NFPA-805 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the AHJ. RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staffs review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

LAR AttachmentS, Table S-2 presents a modification (i.e., MOD-3) to upgrade the reactor coolant pump (RCP) seals. The LAR indicates that credit is taken in the FPRA for this modification. Given recent concerns about the operation of the new Westinghouse RCP shutdown seals, the risk reduction credit that might be taken in this application for upgraded RCP seals may be optimistic. Describe the RCP seal upgrade and discuss the credit taken in the FPRA for the upgrade.

NextEra Response NextEra committed in PBNP LAR 271 dated June 26, 2013 (Reference 1, Enclosure, Attachment S, Table S-2, Item MOD-3) to install the Westinghouse Electric Company (WEC)

SHIELD"' Passive Thermal Shutdown Seal (SDS) into the existing WEC reactor coolant pump (RCP) seals of each Unit to provide a more controllable leak rate, if cooling flow is lost to the RCP seals. Additional information was provided in the PBNP LAR 271 Supplement dated September 16, 2013. (Reference 3)

NextEra is tracking the Westinghouse redesign, qualification testing, and NRC review of their SDS. Alternatives to the SDS were also evaluated as contingency plans in the event that the SDS redesign is not acceptable.

The most recent qualification testing of the redesigned SDS has been documented in WEC report PWROG-14001-P/NP, "PRA Model for the Generation Ill Westinghouse Shutdown Seal,"

PA-RMSC-0499R2, which has been submitted to the NRC but has not yet been approved. This report provides the basis for the SDS credit in the PRA and will be applied to the integrated results in PRA RAI 3a to be submitted in the 120 Day RAI Response. If the report is not approved by the NRC or requires revision, the PRA used in support of the NFPA 805 PRA will be adjusted as necessary.

The major difference from the existing seals is the existence of a polymer ring and a piston ring which prevent a RCP seal LOCA when the seal cooling and injection have failed and the RCPs have stopped running. In this scenario, the SDS seals will reduce leakage to a nominal value of 1 gpm. For those cases where the RCPs are stopped early in the transient, e.g., a loss of offsite power, the Westinghouse SDS greatly reduces the risk of a seal LOCA compared to the seals which are currently installed. The table below provides a comparison of the original WEC RCP seals, the SDS design used for LAR 271 (Reference 1) and data from the most recent SDS design being used in the integrated response to PRA RAI 3a, which will be submitted in the 120 Day RAI Response.

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Feature Original W License Amendment Generation Ill Seals Seals Request 271 (PWROG-14001-P)

Time to trip RCPs to 13 minutes 19.2 minutes 20 minutes avoid seal LOCA given loss of seal injection and cooling Leakage (gpm/pump) 21 gpm/pump 1 gpm/pump (7 .46E-3 1 gpm/pump (9.5E-3 for a loss of seal (79%}, 182 probability of probability of leakage cooling and injection gpm/ pump shutdown seal failure) (Actuation Fails 8.3E-3, Seal with RCPs tripped in (20%), 76 Fails 1.2E-3))*

time gpm/pump (1%), 480 gpm/pump

(.25%)

Leakage (gpm/pump) 480 gpm/pump 480 gpm/pump 480 gpm/pump**

for a loss of seal cooling and injection with RCPs not tripped in time**

• Probability of SDS failure 1s mdependent of RCP tnp

• • 480 gpm/pump is the maximum leakage possible PRA RA114- Use of Incipient Detection NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805.

LAR Attachment H, Table H-1 indicates that FAQ 08-0046, "Incipient Fire Detection Systems",

(ADAMS Accession No. ML093220426) was utilized in the submittal, yet there is no additional evidence that incipient detection was credited in the FPRA. If incipient detection was credited in the FPRA, identify the type of incipient system (in cabinet or area-wide), where it is installed, and a discussion of the credit.

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NextEra Response Incipient detection was not credited in the development of the PBN FPRA. Attachment H of the LAR (Reference 1) contains the approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been used to clarify the guidance in Regulatory Guide (RG) 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of the LAR. See markup of LAR Attachment H, Table H-1, Page H-2, which removed FAQ 08-0046, in Attachment A PRA RAI 15

• PRA Treatment of Dependencies between Units 1 and 2 NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC Staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC Staff to complete its review of the proposed method.

Several references are made in the LAR to cross-ties and shared systems, including the disposition to F&O AS-81-01 in the LAR which also states that the PBNP lEPRA is modeled using duplicate logic in Units 1 and 2 to capture shared systems. LAR Attachment ~ Tables W-6 and W-7 show contribution by fire area to CDF, LERF, b. CDF, and b. LERF, but do not appear to show how the risk contribution from fires originating in one unit is incorporated into the risk for the other unit given the possible physical proximity of fire zones and the existence of shared systems. Explain how the risk contribution of fires in one unit is addressed for the other unit due to the physical layout of the units and shared systems. Include identification of locations where fire in one unit can affect components in the other unit, and description of the extent to which systems are shared. If the contribution of fires originating in one unit is not addressed for the other unit, perform this assessment and include as part of the integrated analysis requested for PRA RAJ 3.

NextEra Response Opposite Unit Impacts Each postulated fire in the plant is evaluated to determine the scope of fire damage, either from direct damage to the equipment or damage to cables powering and/or controlling the equipment. With the exception of fires originating in the unit-specific containment buildings, each postulated fire is evaluated twice, once for each unit. This evaluation is performed regardless of whether the ignition source is from a Unit 1 component, a Unit 2 component, or a component that is common to both units. As such, the reported core damage frequency (CDF),

b.CDF, large early release frequency (LERF), and b.LERF values for one unit implicitly include any equipment lost due to a fire originating on the opposite unit.

The logic structure for the each unit's probable risk significance (PRA) model has equipment from the opposite unit in it that could affect its risk. Additionally, test and maintenance unavailability for opposite unit equipment is included in the other unit's PRA model to properly account for the unavailability of that equipment. The set of cables damaged from fires originating in common areas is used in the risk calculations for both units. For control room abandonment fire scenarios, both units are affected and both units are shut down using control room abandonment procedure(s).

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Since the contribution of fires originating in one unit is addressed for the other unit, no further assessment or integrated analysis is needed.

Locations where fire on one unit affects the other unit The only locations where fire on one unit DOES NOT affect the other unit are in the unit-specific containment buildings. For all other fire locations, each postulated fire is evaluated twice, once for each unit.

Shared Systems The following list describes the extent to which systems are shared between the units:

• 120 (Y) VAC, common instrumentation, e.g., CST level

• 125 VDC, common batteries and battery chargers, e.g., D-05 and D-07

• 13.8 KVAC, common power distribution equipment, e.g., H52-21 and H52-31

• 345 KVAC, common power distribution equipment, e.g., switchyard bus sections

• 4.16 KVAC, switchgear supplying common equipment or equipment that could be cross-tied to the opposite unit, e.g., 1A-05, 1A-06, 2A-05, 2A-06

• 480 VAC, busses supplying common equipment or equipment that could be cross-tied to the opposite unit, e.g., 18-03, 18-04, 28-03, 28-04, 8-08, 8-09

• Auxiliary Feedwater, equipment to remove decay heat, e.g., 1P-29, 1P-53, 2P-29, 2P-53, P-38A, P-388

• Component Cooling Water, swing CCW heat exchangers that can be aligned to either unit, e.g., HX-128 and HX-12C

• Emergency Diesel Generator, backup power supply to 4,160 VAC switchgear, e.g., G-01, G-02, G-03, G-04

• Fuel Oil, fuel oil storage and pumps to common equipment such as EDGs, GT, diesel fire pump

• Fire Protection, provides fire suppression capability and backup water supply to the AFW pumps

• Gas Turbine, backup power supply to 13.8 KV system, e.g., G-05

• Instrument Air, provides oil-free compressed dry air for the operation of various plant equipment, e.g., K-2A, K-28

• Main Steam, provides steam supply to the Turbine-Driven AFW pumps, which can be shared between the units

• Service Air, provides oil-free compressed backup air supply to the lA system, e.g., K-3A, K-38

• Service Water System, provides cooling to common lA and SA compressors (one SA compressor is independent of SW), G-01 and G-02 EDGs (other two EDGs are independent of SW), CCW heat exchangers, and suction water source for AFW pumps

• VN81, provides cooling to PA8 battery and inverter rooms containing common 125VDC components, e.g., D-105, D-106, D-107, D-108, D-109

• VNDG, provides cooling to the rooms containing common EDGs Page 18 of 58

PRA RAI17- Clarification of Recovery Action Listing NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4. 1 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

For the fires areas in LAR Attachment C that credit recovery actions, the fire risk summary concludes that "with the implementation of recovery actions in Attachment G the applicable risk, defense-in-depth, and safety margin were satisfied." LAR Attachment G, however, does not indicate whether any or which recovery actions are credited only for defense-in-depth. LAR Attachment C, Table C-3/ists a number of VFDRs "associated with main control room abandonment actions" that LAR Attachment G, foes not identify an associated recovery action (i.e., VFDR A31-08, A31-11, A31-22, A31-25, A31-26, and A31-27). In light of these observations:

a) Identify which recovery actions listed in LAR Attachment G are identified only for "defense-in-depth" and which are credited explicitly in the FPRA for risk reduction.

b) Since there are no PCSs (aside from the MCR), discuss whether all the actions credited in the FPRA required outside the MCR were considered recovery actions per NFPA 805, the guidance described in RG 1. 205, "Risk-Informed, Performance-Based Fire Protection For Existing Light-Water Nuclear Power Plants," and are listed in LAR Attachment G.

NextEra Response All of the actions in Table G-1 , Recovery Actions, of the LAR (Reference 1) are risk reduction actions and none are credited specifically for defense-in-depth. The recovery actions listed in the table are fire procedure related actions that require the treatment of additional risk and therefore have been modeled in the Fire PRA with the results portrayed in Attachment W of the LAR (Reference 1).

As Point Beach has no Primary Control Station, all the actions taken outside of the main control room that provide an available success path based on nuclear safety performance criteria are credited recovery actions and are listed in Attachment G of the LAR (Reference 1) (per the requirements of NFPA 805 and guidance of RG 1.205). Actions associated with a fire in the main control room are identified as Fire Area FA A31 with these actions being credited in the post-transition case of the Fire PRA model.

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PRA RAI 20

• Defense in Depth and Safety Margin NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4.1 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

LAR Section 4.5.2.2 provides a description of how the reviews of the impacts on defense-in-depth and safety margin were conducted. Provide further explanation of the method used to determine when a substantial imbalance between defense-in-depth echelons existed in the Fire Risk Evaluations (FREs), and identify the types of plant improvements made in response to this assessment. Also, provide further discussion of the approach in applying the Nuclear Energy Institute (NEI) 04-02, "Guidance for Implementing A Risk-Informed, Petformance-Based Fire Protection Program Under 10 CFR 50.48(c)," criteria for assessing safety margin in the FREs.

NextEra Response Defense-in-depth was qualitatively evaluated for each fire area in the Fire Risk Evaluation (FRE). The methods and criteria utilized in the FRE, and one example, are provided below.

DEFENSE-IN-DEPTH Consistency with the defense-in-depth philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:

• A reasonable balance is preserved among 10 CFR 50.48(c) defense-in-depth elements (fire prevention, fire detection, fire suppression, mitigation, nuclear safety capability).

• Over-reliance and increased length of time or risk in performing programmatic activities to compensate for weaknesses in plant design is avoided

• Pre-fire nuclear safety system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers). (This should not be construed to mean that more than one safe shutdown/NSCA train must be maintained free of fire damage.)

• Independence of defense-in-depth elements is not degraded

• Defenses against human errors are preserved

• The intent of the General Design Criteria in Appendix A to 10 CFR 50 is maintained Defense-in-Depth Approach A review of the impact of the variances from deterministic requirements (VFDRs) on defense-in-depth shall be performed, regardless of the risk evaluation method used. The review of defense-in-depth is typically qualitative and should address each of the elements with respect to the proposed change.

Evaluate the fire area for the impact of the VFDRs on fire protection defense-in-depth. Fire protection defense-in-depth is achieved when an adequate balance of each of the following three elements (or echelons) is provided:

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• Preventing fires from starting;

• Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage; and

• Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the defense-in-depth requirement is satisfied if the proposed change does not result in a substantial imbalance among these elements. Table 1 contains additional defense-in-depth guidance.

In evaluating defense-in-depth, it may become necessary to identify and specifically consider potential risk significant fire scenarios that result from impacts to VFDRs. A fire scenario, in this regard, is defined as a unique quantification of a fire damage state (which may include severity factors and probability of non-suppression) multiplied by a conditional core damage probabilities (CCDP) or conditional large early release probability (CLERP) to arrive at a core damage frequency (CDF) or large early release frequency (LERF). For purposes of defense-in-depth, "potentially risk significant" fire scenarios could be characterized as follows, for example:

A scenario in which the calculated risk is equal to or greater than 1E-6/year for CDF and/or 1E-7/year for LERF could be characterized as "potentially risk significant."

A scenario in which the calculated risk falls between 1 E-6/ year and 1E-8/year for CDF, or between 1E-7/year and 1E-9/year for LERF, and where DID echelon 1 and 2 attributes are causing a significant reduction in risk, could be characterized as "potentially risk significant" A scenario in which the calculated risk is less than 1E-8/year for CDF and/or 1E-09 for LERF, regardless of reliance on DID echelon 1 and 2 attributes, may be characterized as "potentially not risk significant". These values are considered "potentially not risk significant" based on being two orders of magnitude below the acceptance criteria of RG 1.174 as referenced by RG 1.205, Revision 1.

A scenario with a high consequence (i.e., CCDP>1 E-1) could be considered "potentially risk significant."

Fire protection features and systems relied upon to ensure defense-in-depth should be clearly identified in the assessment (e.g., detection, suppression system, etc.).

Verify that defense-in-depth is maintained by assessing and documenting that the balance is preserved among prevention of core damage, prevention of containment failure, and mitigation of consequences. Regulatory Guide 1.174 provides guidance on maintaining the philosophy of defense-in-depth that is acceptable for NFPA 805 Fire Risk Evaluations.

Each fire area shall be evaluated for the need to incorporate defense-in-depth enhancements to provide assurance that plant performance goals can be achieved and maintained.

Documentation of these defense-in-depth enhancements can be on a fire area basis and/or tied directly to a VFDR disposition, as appropriate.

Provide the results of the defense-in-depth review in a tabular format, such as that shown in the example in Table 9.1. Defense-in-depth attributes shall be evaluated for applicability to NFPA 805, Section 4.2.3 or 4.2.4 (Ch. 3, as required).

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If a defense-in-depth attribute is credited for NSCA deterministic criteria, licensing action or engineering equivalency evaluation then the system/feature should already be considered to form an integral part of defense-in-depth. The parent echelon of the system/feature should then be evaluated against the process and considerations presented in Table 1, to determine if any improvements or changes are necessary, such as to offset a weakness in another echelon.

If the Fire PRA credits any of the fire protection features or a recovery action to improve the risk profile then these attributes or features should already be considered to form an integral part of defense-in-depth. The parent echelon of the system/feature should then be evaluated against the process and considerations presented in Table 1, to determine if any improvements or changes are necessary, such as to offset a weakness in another echelon.

Defense-in-depth attributes that go above and beyond the existing requirement(s) wilh the purpose of bolstering derived weaknesses within the defense-in-depth elements to maintain an overall balance should be designated as a change or improvement necessary for defense-in-depth.

Note- this may or may not involve a physical improvement to the element, but by virtue of including an attribute that was not required for deterministic or risk reasons, defense-in-depth is considered enhanced.

Features or enhancements required for defense-in-depth warrant consideration for inclusion in the monitoring program.

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Table 1 - Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 1: Prevent fires from starting

• Combustible Material Controls Combustible and hot work controls are fundamental elements of defense-in-depth and

• Hot Work Controls as such are always in place. The issue to be considered during the fire risk evaluation is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance.

Considerations include:

• Creating a new Transient Combustible Free Area

• Creating a new Hot Work Restriction Area

• Modifying an existing Transient Combustible Free Area or Hot Work Restriction Area The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine if additional controls should be added.

Review the remaining elements of defense-in-depth to ensure an over-reliance is not placed on programmatic activities for weaknesses in plant design.

Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage

• Detection system Automatic suppression and detection may or may not exist in the fire area of concern. The

• Automatic fire suppression issue to be considered during the fire risk

• Portable fire extinguishers provided for the evaluation is whether installed suppression area and or detection is required for defense-in-

• Hose stations and hydrants provided for depth or whether suppression/detection needs the area to be strengthened to offset a weakness in

• Pre-Fire Plan another echelon thereby providing a reasonable balance.

Considerations include:

Risk Insights:

• If the variance is never affected in a "potentially risk significant" fire scenario, manual suppression capability may be adequate and no additional systems required.

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Table 1 -Considerations for Defense-in-Depth Determination Method of Providing DID Considerations

• If the fire area requires recovery actions, then as a minimum, detection and manual suppression capability are required, and suppression should be considered.

• If a fire area contains neither suppression nor detection and a recovery action is required, consider adding detection and/or suppression.

Firefighting Activities:

• If firefighting activities in the fire area are expected to be challenging (either due to the nature of the fire scenario or accessibility to the fire location) then both suppression and detection may be required Fire Scenarios:

• If fire scenarios credit fire detection or fire suppression systems, then these should be considered to form an integral part of defense-in-depth Page 24 of 58

Table 1 - Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed

• Walls floors ceilings and structural If fires occur and they are not rapidly detected elem~nts are rated or have been evaluated and promptly extinguished, then the third as adequate for the hazard. echelon of defense-in-depth would be relied upon.

• Penetrations in the fire area barrier are rated or have been evaluated as adequate The issue to be considered during the fire risk for the hazard.

evaluation is whether existing separation is

• Supplemental barriers (e.g., ERFBS, cable adequate (or over relied on) and whether tray covers, combustible liquid additional measures (e.g., supplemental dikes/drains, etc.) barriers, fire rated cable, or recovery actions)

• Fire rated cable are required to offset a weakness in another Guidance provided to operations personnel echelon thereby providing a reasonable detailing the required success path(s) balance.

including recovery actions to achieve nuclear Considerations include:

safety performance criteria.

Risk Insights:

• If the variance is never affected in a "potentially risk significant" fire scenario, internal fire area separation may be adequate and no additional reliance on recovery actions necessary.

• If the variance is affected in a risk significant fire scenario, internal fire area separation may not be adequate and reliance on a recovery action, supplemental barrier, or other modification may be necessary.

• If the consequence associated with the variance is considered high (e.g.,

CCDP>1 E-01 or by qualitative SSD assessment) regardless of whether it is in a risk significant fire scenario, a recovery action, supplemental barriers, or other modification should be considered.

• There are known modeling differences between a Fire PRA and nuclear safety capability assessment due to different success criteria, end states, etc.

Although a variance may be associated with a function that is not considered a significant contribution to core damage frequency, the variance may be considered important enough to the NSCA to retain as a recovery action. 1 Page 25 of 58

Table 1 -Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Operations Insights:

• If the sequence to perform a recovery action is particularly challenging then including the action for defense-in-depth may be considered. 2 The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the fires evaluated and the consequence in the area to best determine options for this element of defense-in-depth.

1 An example would be components in the NSCA associated with maintaining natural circulation at a pressurized water reactor that are not modeled explicitly in the Fire PRA since they are not part of a core damage sequence.

2 An example would be a recovery action that is unique in nature, time critical and/or not included in emergency response procedures such that the MCR staff may not be able to quickly recognize and perform the required action.

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Table 9.1 Defense-in-Depth Impact Review for Fire Area A01-A Changes or Required Required to Improvements Method of Providing to Support Support Necessary for Basis/Justification DID Fire PRA? Deterministic Defense-in-Criteria? Depth?

Echelon 1: Prevent fires from starting Combustible Control This element is adequate is implemented in based on no perceived accordance with NP Yes Yes No weakness of, or over-1.9.9, "Transient reliance on, another Combustible Control". echelon of defense-in-Hot Work Control is depth.

implemented in accordance with NP Yes Yes No 1.9.13, "Ignition Control Procedure."

Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Fire Detection System Yes Yes No Automatic detection is Fixed Fire credited in the Yes Yes No performance-based Suppression analysis. Fire fighting Portable Fire activities are not expected Yes Yes No Extinguishers to be challenging, Hose stations and therefore, no change or Yes Yes No hydrants improvement to the installed system is required to maintain Pre-Fire Plan Yes Yes No defense-in-depth.

Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed Walls, floors ceilings The variances are never and structural affected in a risk elements are rated or significant fire scenario Yes Yes No have been evaluated and internal fire area as adequate for the separation is adequate.

hazard. Installed fire wrap is Penetrations in the fire credited in the area barrier are rated performance-based or have been analysis. Recovery actions Yes Yes No are required to achieve the evaluated as adequate for the nuclear safety hazard performance criteria, and Supplemental barriers therefore, form an integral (e.g., ERFBS, cable Yes Yes No part of defense-in-depth.

tray covers)

Fire rated cable No No No Page 27 of 58

Table 9.1 Defense-in-Depth Impact Review for Fire Area A01-A Changes or Required Required to Improvements Method of Providing to Support Support Necessary for Basis/Justification DID Fire PRA? Deterministic Defense-in-Criteria? Depth?

Guidance provided to There are no significant operations personnel modeling differences detailing the required between the Fire PRA and success path(s) nuclear safety capability including recovery Yes Yes No assessment (i.e., due to actions to achieve different success criteria, nuclear safety end states, etc.) that are performance criteria. contributing to reduce core damage frequency.

As a result of the defense-in-depth assessment in each fire area Fire Risk Evaluation, no additional features, changes, or improvements were identified as being required only for defense-in-depth to ensure an adequate balance of DID features is maintained in each fire area.

Systems and features required to support the Fire PRA and support the deterministic criteria are considered to form an integral part of defense-in-depth and therefore, were not identified as required for DID.

SAFETY MARGIN Based on NEI 04-02, the requirements related to safety margins for the change analysis are described for each of the specific analysis types used in support of the fire risk assessment. The specific safety margin evaluation depended on the change set.

The evaluation addresses whether:

Codes and Standards or their alternatives accepted for use by the NRC are met, and safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provide sufficient margin to account for analysis and data uncertainty.

These evaluations can be grouped into categories. These categories are:

1. Fire Modeling

2. Plant System Performance

3. PRA Logic Model

4. Other

1. Fire Modeling Page 28 of 58

If a performance-based approach is used, the margin between the parameters describing the Maximum Expected Fire Scenario (MEFS) and the Limiting Fire Scenario (LFS), and the process of judging the adequacy of that fire modeling margin, is required for the overall safety margin consideration. The level of review to be performed as part of the safety margin treatment involves the integration of that margin with the potential consequences of the upset, or damage, that may occur given the LFS. The acceptability of the fire modeling margin between MEFS and LFS was judged in the context of the potential severity of the resulting plant system impact if an LFS were to occur. An LFS that causes an inter-system loss of coolant accident (ISLOCA) event would tend to demand a higher margin between MEFS and LFS as compared to an event that causes a degradation of long term decay heat removal.

2. Plant System Performance The development of the fire risk assessment may involve the re-examination of plant system performance given the specific demands associated with the postulated fire event. The methods, input parameters, and acceptance criteria used in these analyses was reviewed against that used for the plant design basis events.

This subtask evaluates the plant system performance given the specific demands associated with the postulated fire event. The methods, input parameters, and acceptance criteria utilized in the risk-informed, performance-based analysis was reviewed against the plant design basis events. This evaluation determined if the safety margin established in the plant design basis events is preserved.

3. PRA Logic Model This subtask evaluates results of the Fire PRA model to verify that the safety margins have not changed. The contribution to the CDF and LERF results of components in the cutset results was evaluated to verify that events with high contribution have reasonable failure probabilities for the scenarios of interest. This was particularly important for human error basic events. The results of each risk evaluation were evaluated against the base case fire results to determine that no single event has undue influence on the results of the change analysis. This evaluation demonstrated that the safety margin established in the PRA model is preserved and that the Fire PRA model is sufficient to treat the fire-induced core damage sequences.

4. Other (referred to as Miscellaneous in NEI 04-02)

This category addresses any other analyses not addressed above. The general requirements related to codes and standards, and acceptance criteria, provided above, apply.

Example of a Typical Safety Margin Review as Contained in a Fire Risk Evaluation for a Fire Area with One or More VFDRs In accordance with NEI 04-02, the maintenance of adequate Safety Margin is assessed by the consideration categories of analyses utilized by this Fire Risk Evaluation.

Safety margins are considered to be maintained if:

Codes and Standards or their alternatives accepted for use by the NRC are met.

AND Page 29 of 58

Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met or provide sufficient margin to account for analysis and data uncertainty.

The following summarizes the bases for ensuring the maintenance of safety margins:

The risk-informed, performance-based processes utilized are based upon NFPA 805, 2001 edition, as endorsed by the NRC in 10 CFR 50.48(c).

The Fire Risk Evaluation process is in accordance with NEI 04-02, Revision 2, which is endorsed by the NRC in Regulatory Guide 1.205, Revision 1.

The Fire PRA is developed in accordance with NUREG/CR-6850, which was developed jointly between the NRC and EPRI.

The Fire PRA has undergone an industry peer review in order to ensure the Fire PRA meets the appropriate quality standards of ASMEIANS RA-Sa-2009, "Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications." This peer review also served to evaluate compliance with Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2.

The PBNP internal events PRA model received a formal industry peer review, conducted in accordance with the applicable NEI guidelines. Peer review of the PRA model has been conducted by diverse groups of PRA practitioners from other PWR plants and industry. The PWROG peer review covered all aspects of the PBNP PRA model and administrative processes used to maintain the model. Peer review has generated specific recommendations for model changes, as well as guidance for improvements to processes and methodologies used in the PRA model, and enhancements to the documentation of the model and the administrative procedures used for model updates. The PRA model and administrative requirements are assessed and revised or clarified to address the issues identified through peer reviews.

Fire protection systems and features determined to be required by NFPA 805 Chapter 4 have been confirmed to meet the requirements of NFPA 805 Chapter 3 and their associated referenced codes and listings, or provided with acceptable alternatives using processes accepted for use by the NRC (i.e., FAQ 06-0008, FAQ 06-0004, FAQ 07-0033).

Fire modeling performed in support of the transition has been performed within the Fire PRA utilizing codes and standards developed by industry and NRC staff which have been verified and validated in authoritative publications, such as NUREG 1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications." In general, the fire modeling, where performed in support of the Fire Risk Evaluations, has been performed using conservative methods and input parameters that are based upon NUREG/CR-6850. While this is generally not ideal in the context of best estimate probabilistic risk analysis, it is a pragmatic approach given the current state of knowledge regarding the uncertainties related to the application of the fire modeling tools and associated input parameters for specific plant configurations.

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In accordance with the requirements of 10 CFR 50.48(c)(iii), the Fire PRA results, including cutsets for the scenarios of concern, have been reviewed, and it was verified that the results presented above do not rely solely on feed and bleed as the fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability for this fire area.

The "detailed risk evaluation" is used during transition (NEI 04-02, Section 5.3.4.3); therefore, MEFS/LFS is not analyzed separately from the Fire PRA results.

PRA RAI 21 -Implementation Item Impact on Risk Estimates NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4.1 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

Implementation item IMP-142 in LAR AttachmentS, Table S-3 states that FPRA model will be updated after modifications are complete, however, it does not indicate that an update to the FPRA will occur following completion of other implementation items and does not identify a plan of action if RG 1. 174 guidelines are exceeded. Describe how the FPRA model will be updated after all implementation items are complete, and if RG 1.174 guidelines are exceeded.

NextEra Response The FPRA model will be updated per the NextEra Fleet procedure on "PRA Configuration Control and Model Maintenance." This procedure requires periodic data analysis and model updates to be performed at an interval which should not exceed 5 years.

This procedure is being revised to include a requirement regarding conformance to RG 1.174.

This addition will clearly state that if the criteria in RG 1.174 is exceeded, the change cannot be implemented. Therefore, this condition will require a look at alternatives that include resubmittal of the LAR or modifications to bring the result within acceptable RG 1.174 limits.

A markup to the LAR Attachment 8, Table 8-3, IMP-142 is provided in Attachment B.

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PRA RAI 22 - Model Changes and Focused Scope Reviews after the Full Peer Review NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. RG 1.200 describes a peer review process utilizing an associated ASMEIANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established.

The disposition to F&O HRA-02-01 states that a dependency analysis for fire event HFEs did not exist at the time of the FPRA peer review, but was performed after the peer review and is documented in the licensee's analysis. It does not appear to the NRC staff that a peer review was performed on this model upgrade.

a) Identify any changes made to the lEPRA or FPRA since the last full -scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, "Standard for Levei1/Large Early Release Frequency for Nuclear Power Plant Applications," as endorsed by RG 1.200, 'i!\n Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities."

b) If any changes are characterized as a PRA upgrade, indicate if a focused-scope peer review was performed for these changes consistent with the guidance in ASMEIANSRA-Sa-2009, as endorsed by RG 1. 200, and describe any findings from that focused-scope peer review and the resolution of these findings.

c) If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to address this issue.

NextEra Response a) Changes to the Fire PRA model were made in the areas of Technical Elements Fire Risk Quantification (FQ) and Fire Scenario Selection and Analysis (FSS), since the full-scope FPRA peer review was performed in June 2011. Since these changes were considered upgrades, focused-scope peer reviews were performed for these elements in May 2013 (FSS) and June 2013 (FQ). The FPRA dependency analysis is not considered a PRA upgrade because the methodology used for the FPRA is the same as the methodology used for the lEPRA, which was included in the lEPRA peer reviews. Changes to the lEPRA were made in the area of the resolution of findings, but none of these was considered PRA upgrades.

b) If a PRA change is an upgrade, a peer review is required per the NextEra Fleet procedure for PRA Configuration Control and Model Maintenance.

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Focused-scope peer reviews were performed in May 2013 for FSS element and in June 2013 for FQ element of the FPRA. Changes to the FPRA model subsequent to the focused-scope peer reviews were associated with resolution of F&Os. The resolution of these findings did not constitute a PRA upgrade and are included in the Attachment V of the LAR 271 (Reference 1).

For the lEPRA, a limited scope peer review was performed in October 2011 (resolution of findings). The resolution of these findings did not constitute a PRA upgrade. Changes to the lEPRA model subsequent to the focused-scope peer review were associated with resolution of F&Os and are included in the Attachment U of the LAR 271 (Reference 1).

c) Focused-scope peer reviews have been performed for all FPRA and lEPRA changes characterized as PRA upgrades.

• For the lEPRA, a focused-scope peer review was performed in August 2011 {resolution of findings). The resolution of these findings did not constitute a PRA upgrade. Changes to the lEPRA model subsequent to the focused-scope peer review were associated with resolution of F&Os and are included in the attachment U of the LAR 271 (Reference 1).

Focused-scoped peer reviews have been performed for all lEPRA and FPRA updates identified as "PRA upgrades" per ASME/ANS-RA-Sa-2009, "Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," as endorsed by RG 1.200, "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities."

PRA RAI 23 - Modification Credit in Internal Events PRA NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4.1 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1. 174 provides quantitative guidelines on core damage frequency, large early release frequency, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff's review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

LAR Attachment W, Table W-1 reports low CDFs (3.1 E-6/year for each unit) for the internal events hazard. Clarify whether this CDF reflects the risk reduction due to non-VFDR modifications identified as credited in the FPRA in LAR AttachmentS, Tables S-1 and S-2.

NextEra Response The plant modifications listed in Table S-1 of the LAR (Reference 1) have already been completed and installed in the plant. These modifications were not explicitly performed for NFPA 805, although they were installed while the analysis for NFPA 805 was ongoing. All of these modifications are included in the internal events CDF and LERF values reported in Table W-1.

The plant modifications listed in Table S-2 are not yet installed in the plant. The committed plant modifications that were included in the internal events CDF and LERF values reported in Table W-1 are as follows:

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• Replace TDAFWPs with self-cooled models. Note that this modification was already completed for 2P-29 during the spring 2014 Unit 2 refueling outage. (Table S-2, Item EC 272527, EC 272529)

• Cross tie TDAFWP steam supplies and pump discharge(s) to allow opposite Unit support. (Table S-2, Item MOD-2)

• The RCP Seals will be upgraded. WCAPs 17100-P-A and 17541-P. (Table S-2, Item MOD-3)

• Add additional power inputs to B-08 and to B-09 to increase the current load capability. Power to come from tie line independent of switchyard. (Table S-2, Item MOD-4)

• Provide redundant power supply to P-38s from B-08, B-09. (Table S-2, Item MOD-5)

Two additional plant modifications listed in Table S-2 would lower the risk from internal events, but were not included in the internal events values reported in Table W-1. These two plant modifications are as follows:

• Restore N2 supply to primary PORVs and verify supply is adequately sized to support PRA success criteria of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Table S-2, Item MOD-6)

• Reduce dependence on instrument air for P-38 AOVs by providing accumulators with 24-hour air supply. (Table S-2, MOD-23}

All of the other plant modifications listed in Table S-2 are being done to lower risk in the fire PRA and would not benefit the internal events PRA values.

PRA RAJ 24 ~ Smoke Damage NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, Rev. 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described in guidance documents require additional justification to allow the NRC staff to complete its review of the proposed method.

The licensee's analysis indicates that, although qualitative analyses are performed of smoke damage, no fire modeling for smoke effects is performed. It is not clear how smoke effects were addressed in the FPRA. Explain how effects of smoke on equipment were evaluated using applicable guidance (i.e.,

Appendix T of NUREG/CR-6850) or some other method.

NextEra Response Smoke damage to equipment was evaluated using the guidance provided in NUREG/CR-6850, Appendix T. Consistent with Section T.3.1, short-term smoke damage was only assumed to result from a severe smoke exposure condition. Therefore, components housed in the same electrical panel as the fire source, or in an electrical panel directly connected via an open bus duct, were assumed damaged by smoke, unless a specific installation feature, such as the features identified in NUREG/CR-6850, Section T.3.1, precludes such damage. Any smoke damage is detailed in Attachment 1, Detailed Fire Modeling Workbook of each individual fire compartment.

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Fire Modeling (FM) RAI 02 ASME/ANS Standard RA-Sa-2009, "Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications," Part 4, requires damage thresholds be established to support the FPRA. Thermal impact(s) must be considered in determining the potential for thermal damage of SSCs and appropriate temperature and critical heat flux criteria must be used in the analysis.

a) Describe how the installed cabling in the power block was characterized, specifically with regard to the critical damage threshold temperatures and critical heat fluxes for thermoset and thermoplastic cables as described in NUREG/CR-6850.

b) It appears that, for covered trays, a damage delay time is assumed based on NUREG/CR-6850, Vol.

2, Appendix Q, Section 0 .2.2. However, the delay time recommended in this section of NUREG/CR-6850 should only be used for qualified cable.

Confirm that all cables for which the delay time was assumed are qualified.

NextEra Response a) Cables in conduit and raceways have been conservatively analyzed as thermoplastic targets in all fire compartments at Point Beach. In accordance with NUREG/CR-6850, Appendix H, Table H-1, the following damage criteria for temperature and radiant heat flux were used for thermoplastic targets:

• Critical Temperature: 205°C (400°F)

• Critical Heat Flux: 6 kW/m 2 (0.5 BTU/fFs) b) Cables in conduit and raceways have been conservatively analyzed as thermoplastic targets in all fire compartments at Point Beach and delay time is assumed as follows:

Cable trays provided with solid bottom covers were credited to delay, by 4 minutes, damage to and ignition of thermoplastic cables, based on the test results from NUREG/CR-0381, "A Preliminary Report on Fire Protection Research Program Fire Barriers and Fire Retardant Coatings Tests." No tests that were performed on PVC (i.e., unqualified) cable with a solid bottom tray and no coating had a time to electrical short or a time to ignition that was less than 4 minutes.

Per Detailed Fire Modeling Report R2168-001-318, fire growth and propagation was not postulated for any fully enclosed cable trays in the Cable Spreading Room (Fire Compartment 318). These cable trays are robustly enclosed on all sides with heavy gauge steel and 1'2 Kaowool insulation is provided on top of the cables below the top cover. Therefore, the barriers are credited to delay cable damage until after automatic suppression activation, which is slightly greater than the 4 minute delay credited for cable trays with solid bottom covers only. Attachment 5 of R2168-001-318 provides additional justification to credit a 6 minute delay in cable damage for the fully enclosed trays with Kaowool in the Cable Spreading Room.

FM RAI 03 NFPA 805, Section 2. 7.3.2, states that each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.

LAR Section 4.5.1.2 states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). Reference is made toLAR Attachment J, for a discussion of the verification and validation (V& V) of the fire models that were used. Furthermore LAR Page 35 of 58

Section 4. 7.3 states that "calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2. 7.3.2 of NFPA 805."

Regarding the V& V of fire models:

a) LAR Attachment J states that the smoke detection actuation correlation (Method of Heskestad and Delichatsios) has been applied within the validated range reported in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications." However the latter reports a validation range only for Alpert's ceiling jet temperatures correlation. Provide technical details to demonstrate that the temperature to smoke density correlation has been applied within the validated range, or to justify the application of the correlation outside the validated range reported in the V& V basis documents.

b) For any tool or method identified in the response to FM RAJ 01 (a) above, provide the V&V basis if not already explicitly provided in the LAR (for example in LAR Attachment J).

NextEra Response a) The Heskestad and Delichatsios Smoke Detection Actuation Correlation is based upon the ceiling jet temperature predicted by Alpert's Ceiling Jet Correlation; therefore, the normalized parameters for the ceiling jet correlation are applicable. The normalized parameter that applies to the Alpert's ceiling jet correlation is the ceiling jet radial distance relative to the ceiling height, and the validation range is 1.2-1.7. The Heskestad and Delichatsios Smoke Detection Actuation Correlation using Alpert's Ceiling Jet Correlation was performed within the validated range reported in NUREG-1824, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, therefore, the correlation was appropriately applied.

In addition to being applied within the validation range for Alpert's ceiling jet, the smoke detection correlation was applied to fuels, configurations, and environmental conditions consistent with those described in Chapter 4-1 of the Society of Fire Protection Engineers (SFPE) Handbook and NUREG-1805, Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Program. The correlation was also applied within the limitations described in these publications.

Heskestad and Delichatsios correlated a smoke temperature change of 1ooc (18°F) based upon typical fire fuels. The materials tested to develop the Heskestad and Delichatsios smoke detector correlation are representative of the fuels modeled for smoke detector activation. The tested materials include various plastics, foams, and paper, possessing smoke properties similar to the fires modeled at PBNP.

Additionally, per EPM Procedure EPM-DP-FP-001, Detailed Fire Modeling, when implementing the Heskestad and Delichatsios Smoke Detection Actuation Correlation (i.e. FDT1 0), the 1ooc (18°F) ceiling jet temperature rise from ambient temperature is preserved by adjusting the activation temperature of the smoke detector accordingly.

b) The solid flame radiation model (method of Shokri and Beyler) is the only tool or method identified in the expected response to FM RAI 01 (a). The model was used to calculate the radiative heat flux from a fire in the Main Control Room Analysis (P2091-2700-01) and Structural Steel Analysis (P2091-2920-02).

The use of this correlation has been verified and validated in Appendix E of Report R2168-1 0038-001.

Section 4.5.1.2 and Attachment J of the LAR will be revised to include the solid flame radiation model and a markup of the LAR is attached to the response for FM RAI 01 (a) to be submitted in the 90 Day RAI Response.

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FM RAI 05 LAR Section 4.5.1.2 states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). The NRC staff notes that this requires that qualified fire modeling and PRA personnel work together. Furthermore, LAR Section 4. 7.3 states that post transition, for personnel performing fire modeling or Fire PRA development and evaluation, NextEra will develop and maintain qualification requirements for individuals assigned various tasks. Position specific guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2. 7. 3. 4 to perform assigned work.

Regarding qualifications of users of engineering analyses and numerical models (i.e., fire modeling techniques):

a) Describe the requirements to qualify personnel for performing fire modeling calculations in the NFPA 805 transition.

b) Describe the process for ensuring that fire modeling personnel have the appropriate qualifications, not only before the transition but also during and following the transition.

c) When fire modeling is performed in support of the FPRA, describe how proper communication between the fire modeling and FPRA personnel is ensured.

NextEra Response a) Fire modeling calculations have been, and will be, performed by engineers who meet the qualification requirements of Section 2.7.3.4 of NFPA 805 (2001 ).

Fire modeling to support the LAR and Fire PRA development was performed by contract personnel using their companies' procedures and their Quality Assurance programs. These procedures require that project personnel assigned to each task have the proper experience and training to perform the work as determined by their company processes.

These contractor personnel were chosen based on their experience and expertise in fire modeling. The qualifications needed to perform fire modeling-related tasks depends in part on the specific role of the personnel. Appropriate qualifications for contractor personnel using, applying, and approving fire modeling tools include required reading on fire modeling Project Instructions, relevant industry methodology and/or guidance documents such as NUREG/CR-6850, NUREG-1934, NUREG-1805, and applicable fire modeling software user's guide documents. Other requirements include training and/or mentoring in Fire Growth Analysis, Zone of Influence (ZOI) calculations, and Fire Modeling Tools.

Qualification requirements also involve a demonstration of comprehension and proficiency in fire modeling.

b) The qualification requirements to perform other fire modeling related tasks depend in part on the personnel's specific assigned role. Some sub-tasks of fire modeling, may be assigned to other staff with experience and skill set commensurate with the task Page 37 of 58

During the NFPA 805 transition phase (i.e., LAR submittal to receipt of the NRC Safety Evaluation),

NextEra will continue to utilize qualified personnel. The NextEra Fleet will not be qualifying personnel at the sites for the purpose of performing fire modeling calculations. Rather, the sites will employ the services of qualified vendors to meet the requirements of NFPA 805 Section 2.7.3, Quality. Point Beach Fire Protection personnel will be trained during the implementation phase to provide an overview of the NFPA 805-2001 Edition Performance-Based Standard for Fire Protection Fire Modeling theories and the requirements of the Point Beach specific Fire Model calculations. The Fire Protection and Fire PRA engineers will have enough understanding of the fire modeling requirements to determine if the modeling calculation is affected. Appropriate qualification guides will be developed under Implementation items IMP-135 (Reference 1, Table S-3).

Fire modeling to support the LAR and Fire PRA development was performed by contract personnel using their companies' procedures and their Quality Assurance programs. These procedures require that project personnel assigned to each task have the proper experience and training to perform the work as determined by their company processes.

These contractor personnel were chosen based on their experience and expertise in fire modeling. The qualifications needed to perform fire modeling-related tasks depends in part on the specific role of the personnel. Appropriate qualifications for contractor personnel using, applying, and approving fire modeling tools include required reading on fire modeling Project Instructions, relevant industry methodology and/or guidance documents such as NUREG/CR-6850, NUREG-1934, NUREG-1805, and applicable fire modeling software user's guide documents. Other requirements include training and/or mentoring in Fire Growth Analysis, ZOI calculations, and Fire Modeling Tools. Qualification requirements also involve a demonstration of comprehension and proficiency in fire modeling.

Point Beach will not be qualifying individuals for the task of performing fire modeling calculations.

Qualifications of the Fire Protection Engineer will be updated during implementation to address the additional knowledge requirements associated with NFPA 805.

c) Fire modeling performed to support the Fire PRA was developed into approved calculations, which provided input into the Fire PRA. These documents are controlled under the Point Beach design and configuration management processes. During the development of the Fire PRA, fire modelers maintained frequent communication with the PRA engineers. This was realized through the project team meetings, cutset reviews, and schedule updates during the development of the Fire PRA.

Implementation item IMP-135 will provide position specific training to ensure that the Point Beach personnel continue the vital communication aspect of the process.

FM RAI 06 LAR Section 4. 7.3, states that uncertainty analyses were performed as required by 2. 7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and Fire PRA development.

Regarding the uncertainty analysis for fire modeling:

a) Describe how the uncertainty associated with the fire model input parameters (compartment geometry, radiative fraction, thermophysical properties, etc.) was addressed for this application and accounted for in the analyses.

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b) The NRC staff notes that cabinets and cable trays reduce the effective volume of a comparlment, but

. also act as a heat sink and that it is not clear how ignoring these contents affects the HGL temperature.

Explain how the corresponding uncerlainties were accounted for and discuss the HGL temperatures resulting from ignoring the comparlment contents and whether they are conservative.

NextEra Response a) Fire modeling was performed within the Fire PRA, utilizing codes and standards developed by industry and NRC staff and that were verified and validated in authoritative publications such as NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications." In general, the fire modeling was performed using conservative methods and input parameters that were based upon NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities." This approach was used based upon the current state of knowledge regarding the uncertainties related to the application of the fire modeling tools and associated input parameters for specific plant configurations. A discussion of uncertainties associated with detailed fire modeling is summarized below.

The detailed fire modeling task developed a probabilistic output in the form of target failure probabilities that were subject to both aleatory (statistical) and epistemic (systematic) uncertainties.

Appendix V of NUREG/CR-6850 recommends that to the extent possible, modeling parameters should be expressed as probability distributions and propagated through the analysis to arrive at target failure probability distributions. These distributions should be based on the variation of experimental results as well as the analyst's judgment. To the extent possible, more than one fire model can be applied and probabilities assigned to the outcome which describe the degree of belief that each model is the correct one. .

• The propagation of fire for each non-screened fire source has been described by a fire model (represented by a fire growth event tree) which addresses the specific characteristics of the source and the configuration of secondary combustibles.

Aleatory uncertainties identified within the fire modeling parameters include:

• Detector response reliability and availability

• Automatic suppression system reliability and availability

• Manual suppression reliability with respect to time available Epistemic uncertainties which impact the zone of influence and time to damage range include:

• Heat release rates (peak HRR, time to reach peak, steady burning time, decay time)

• Number of cabinet cable bundles

• Ignition source fire diameter

• Room ventilation conditions

• Fire growth assumptions (cable tray empirical rule set, barrier delay)

• Cable fire spread characteristics for horizontal and vertical trays

• Transient fires (peak HRR, time to reach peak, location factor, detection time)

• Oil fires (spill assumptions)

• Assumed target location

• Target damage threshold criteria Page 39 of 58

• Manual detection time

• Mean prompt suppression rate

• Manual suppression rate

• Welding and cutting target damage set

• Transient target impacts Due to the uncertainty with each of these parameters, the fire modeling task has selected conservative values for each to provide safety margin. Per NEI 04-02, there is no clear definition of an adequate safety margin. However, the safety margin should be sufficient to bound the uncertainty within a particular calculation or application. Each Detailed Fire Modeling Report, the Compartment Analysis Notebook, and the Detailed Fire Modeling in Selected Point Beach Nuclear Plant Fire Zones, provides a list of items that were modeled conservatively and that provide safety margin for the compartment analyzed. Some examples include the following items:

• The majority of fire scenarios involving electrical equipment (including the electrical split fraction of pump fires) utilize the 98th percentile HRR for the severity factor calculated out to the nearest FPRA target. This was considered conservative.

• The fire elevation in most cases was at the top of the cabinet or pump body. For Containment Building Scenarios, the base of the fire is assumed to be located 1.8m (6ft) above the floor for all fixed sources. This was considered conservative, because the combustion process will occur where the fuel mixes with oxygen, which is not always at the top of the ignition source.

• A radiative fraction of at least 0.3 was utilized in most cases. This value is recommended by the SFPE and any greater value introduced additional conservatism. The Detailed Fire Modeling Workbooks utilized a radiative fraction of 0.4, increasing conservatism by 33%.

• The convective heat release rate fraction utilized was 0.7. The normally recommended value is between 0.6 and 0.65, and thus the use of 0.7 was conservative.

• For transient fire impacts, a large bounding transient zone assumes all targets within its Zone of Influence (ZOI) were affected by a fire. Time to damage is calculated based on the most severe (closest) target. This was considered conservative, because a transient fire would actually have a much smaller zone of influence and varying damage times. This approach was implemented to minimize the multitude of transient scenarios to be analyzed.

• For hot gas layer calculations, no equipment or structural steel was credited as a heat sink, because the closed-form correlations used do not account for heat loss to these items.

• Not every cable tray was filled to capacity. In most cases fire modeling assumed all cable trays were filled to capacity, which provided a conservative estimate of the contribution of cable insulation to the fire and the corresponding time to damage. In some instances additional information on cable loading was used to reduce this capacity in the model, while still using conservative estimates.

• As the fire propagated to secondary combustibles, the fire was conservatively modeled as one single fire using the fire modeling closed-form correlations. The resulting plume temperature estimates used in this analysis were therefore also conservative, because in actuality, the fire would be distributed over a large surface area, and would be less severe at the target location.

• For most scenarios, target damage was assumed to occur when the exposure environment met or exceeded the damage threshold. No additional time delay due to thermal response was allowed.

• The fire elevation for transient fires was assumed to be 2 feet in the Detailed Fire Modeling Reports and Detailed Fire Modeling in Selected Point Beach Nuclear Plant Fire Zones report.

The Compartment Analysis Notebook assumed a 1 meter elevation. Both approaches are Page 40 of 58

considered conservative since most transient fires are expected to be below this height (e.g.,

at floor level).

• Oil fires were analyzed in the Detailed Fire Modeling Reports and Compartment Analysis Notebook as both unconfined and confined spills with 20-minute durations. While unconfined spills resulted in large heat release rates, they usually burn for seconds, not minutes.

However, all the oil fires in the aforementioned reports have been conservatively analyzed for a 20-minute burn time to account for the uncertainty in the oil spill size. The Detailed Fire Modeling in Selected Point Beach Nuclear Plant Fire Zones report conservatively assumes whole room damage or damage to an entire elevation for oil fires depending on the anticipated spill size.

• High energy arcing fault scenarios were conservatively assumed to be at peak fire intensity for 20-minutes from time zero (ignition), even though the initial arcing fault is expected to consume the contents of the cabinet and burn for only a few minutes.

• For many fire scenarios, fire brigade intervention was not credited prior to 85 minutes. A review of the Fire Brigade drills indicated that typical manual suppression times can be expected to be much less than 85 minutes (i.e., 20 minutes).

b) For uncertainties associated with omitting the compartment contents from hot gas layer (HGL) calculations, there were several areas of conservatism that mitigated the effect of increased volume. The following assumptions and justifications were utilized within the fire modeling which lead to conservatism or reduced the impact of omitting the contents of a compartment in the fire modeling analysis:

• Including equipment and cable trays in HGL calculations provides a large heat sink in the compartment, which would result in lower HGL temperatures. Removing these heat sinks conservatively increases HGL temperatures.

• No heat transfer through fire doors or dampers was considered in the HGL temperature calculations. The material properties of concrete were applied to all exterior boundaries of the fire compartment. This provides conservatism since fire doors and dampers are typically thinner and made of more conductive materials than concrete walls. Omitting these fire doors and dampers from HGL temperature calculations reduces heat loss from the room, conservatively increasing HGL temperatures.

• Although obstructions within the room could reduce the effective volume when analyzing HGL temperatures, many of these obstructions (e.g., electrical cabinets, transformers) are not completely solid. Electrical cabinets, for example, are generally not entirely full of electrical equipment and can contain significant amounts of empty space. Therefore, these obstructions do not reduce the compartment volume as much as their physical dimensions indicate. This reduces the impact of obstruction volume on HGL temperature calculations.

• The volume of some fire compartments was reduced in the fire modeling analysis to meet the validation range for compartment aspect ratio. For fire compartments having an aspect ratio outside the validated range, where detailed fire modeling was performed, and whole room damage was not postulated, the height, length, or width of the fire compartment was "shortened" to values that fall within the validation range. Shortening the dimensions of the fire compartment decreases the overall volume of the compartment and creates more severe environmental conditions. The reduction in volume in these compartments conservatively bounds the obstructions that were not considered.

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In addition to the conservatisms specifically related to compartment volume and heat transfer, many of the conservatisms built into the safety margin for all models help alleviate this uncertainty as well. The safety margin is discussed in detail in the response to PBNP RAI FM 06.a and includes conservative inputs summarized as follows:

• Use of 981h percentile heat release rate (HRR)

• Conservative fire elevations

• Increased Radiative/Convective fractions

• Conservative cable tray HRRs

• Propagation to secondary combustibles modeled as a single fire

• No damage delay due to thermal response time

• Conservative oil fire duration and severity

• Conservative high energy arcing fault (HEAF) HRR and duration

• Conservative fire brigade response times.

Safe Shutdown Analysis (SSA) RAI 01 The NRC staff noted that some implementation items and modifications are referenced in the LAR, but not in sufficient detail to determine what particular implementation item or modification relates to the proposed change. For the items listed below, provide the following:

a) Identify the specific modifications in LAR AttachmentS that correlate with elements 3.5.2.4 and 3.5.2.5 of LAR Attachment B, Table B-2. The alignment basis in LAR Attachment B, Table B-2, for elements 3.5.2.4 and 3.5.2.5 refer toLAR AttachmentS, Table S-2 for modifications associated with circuit coordination and common enclosure criteria. There are several modifications in LAR Attachment S associated with circuits, breakers, and fuses.

b) Identify the implementation item(s) that address the revision to the training program and drill procedures to incorporate the feasibility evaluation results. LAR Attachment G, under the heading, "Results of Step 4," describes implementation items resulting from the feasibility evaluation including revision to the training program and revision to the drill development procedure and states these items are included in LAR Attachment S.

c) In LAR Attachment C, the Fire Risk Summary for Fire Areas A01-B/46, A23N, and A36, states, in part, that with the proposed cable protection in Attachment S, the applicable risk, defense-in-depth, and safety margin criteria were satisfied. There were no VFDR dispositions identified in these fire areas that describe modifications.

i. Confirm the modifications referenced in the Fire Risk Summary for the individual fire areas are not associated with a VFDR disposition ii. Identify the specific modification(s) item in LAR Attachment S that is/are associated with the risk summaries in Attachment C for these areas.

NextEra Response a) LAR Attachment B, Element 3.5.2.4, "Circuit Failures Due to Inadequate Circuit Coordination",

indicates four (4) references to AttachmentS (Reference 1).

The specific modifications in LAR Attachment S that correspond with these references are as follows:

• MOD-26-1 for 480V Motor Control Centers- MCC B-21 Coordination (Page B-85)

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• MOD-26-3 for 120VAC Distribution Panel Coordination (Page B-85)

• MOD-26-3 for 120VAC Safety Related Instrument Bus Coordination (Page B-86)

• MOD-26-3 for 120VAC Branch Circuit Coordination (Page B-86)

LAR, Attachment B, Pages B-85 and B-86 are revised to add the applicable MOD number to each AttachmentS statement and AttachmentS, Page S-16 and S-17 are revised to clarify the specific MOD-26 applicable to these concerns. A markup of applicable LAR pages is provided in Attachment C.

LAR Attachment B, Element 3.5.2.5, "Circuit Failures Due to Common Enclosure Concerns", indicates three (3) references to Attachment S.

The specific modifications in LAR Attachment S that correspond with these references are as follows:

• MOD-17 and MOD-30 for 13kV & 4kV Common Enclosure Concerns (Page B-87)

• MOD-24 for cables not protected for overload (Page B-88)

• MOD-26-1 and MOD-26-3 for 480V MCCs and Power Panels and 208/120 Lighting Panel cable protection (Page B-88)

LAR, Attachment B, Page B-87 and B-88 are revised to add the applicable MOD number to each AttachmentS statement. A markup of applicable LAR pages is attached to this RAI response.

LAR Attachment G, in "Results of Step 4" states that "implementation items resulting from the feasibility evaluation are included in the corrective action program. These items include:

• Development/revision of procedures

• Revisions to the Training Program to reflect procedure changes

• Revision to the drill development procedure

• These items are included in Table S-3."

b) Several Table S-3 Implementation Items are related to the Feasibility Evaluation, to include:

IMP-135 (Pg S-27) Fire protection program documents will be updated and training will be provided as necessary. This includes fire protection design basis document, system-level design basis documents and procedures, the Fire Protection Evaluation Report (FPER), the Fire Hazards Analysis Report

{FHAR), the Safe Shutdown Analysis Report (SSAR), post transition change process (including Fire PRA updates), and qualification training. This is being tracked by NAMS Action Request 1882226.

IMP-143 (pg S-29) A confirmatory demonstration (field validation walk-through) of the 4.2.1.3 and Attachment G feasibility for the credited NFPA 805 Recovery Actions (RA) will be performed. This will include field validation 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 RA 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.

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This is being tracked by NAMS Action Request 1882226.

Note that IMP-152 is updated to:

IMP-152 (pg. S-30) Procedure FOP 1.2, "Potential Fire Affected Safe Shutdown Components," will be revised from guideline format to utilize a procedure-type format; and requisite training will be performed for the revised procedure once formally issued. The feasibility for each fire-specific safe and stable action, including a formal walk-through and a timing evaluation, will be evaluated and documented. This is being tracked by NAMS Action Request 1882226.

c) The RAI inquires about three (3) specific Fire Areas.

• A01-B/46 - Proposed cable protection referenced in the Fire Risk Summary for Fire Area A01-B/46 (MOD-18) is not related to VFDR dispositions. Per the Fire Risk Evaluation, it is a risk reduction modification. The Fire Risk Summary for Area A01-B/46 (pg C-53) is revised to add reference to MOD-18 "proposed cable protection in AttachmentS Table S-2 MOD-18."

• A23N - Proposed cable protection referenced in the Fire Risk Summary for Fire Area A23N (MOD-12) is not related to VFDR dispositions. Per the Fire Risk Evaluation, it is a risk reduction modification.

• MOD-17 (LAR Table S-2 pg S-10) protects multiple cables in various Fire Areas to preserve DC control power to multiple buses to ensure the ability of 4160V breakers to trip on an overcurrent (OCT) condition (see detailed response and discussion of OCT modifications in the response to RAI SSA-05, which will be provided in the 120 day RAI response).* Some of the cables to be protected are in Area A23N.

Cable ID Assoc VFDR Remarks ZFD0206A A23N-27 VFDR disposition does not mention cable protection since it does not fully resolve VFDR.

ZFD1402A1/A2 A23N-10 VFDR disposition does not mention cable protection since it does not fully resolve VFDR.

ZFD0203A A23N-31 New VFDR as a result of EC-261 022 and further evaluation under RAI SSA-05.

ZFD0208A A23N-31 New VFDR as a result of EC-261022 and further evaluation under RAI SSA-05.

The Fire Risk Summary in LAR Attachment C (Reference 1) for Area A23N (pg C-264) is revised to add reference to MOD-17 (i.e. "proposed cable protection in AttachmentS Table S-2 MOD-17"). See marked up LAR pages in Attachment C. Since the VFDRs are more closely related to the OCT issue than to their particular impact in A23N, and the VFDRs are not fully resolved by the cable protection, the above A23N VFDR dispositions do not require an update.

A36- Proposed cable protection referenced in the Fire Risk Summary for Fire Area A36 (MOD-8, MOD-

9) are not related to VFDR dispositions. Per the Fire Risk Evaluation, these are risk reduction modifications. MODs 8 and 9 intend to protect RCS pressure instrument cables in Area A36 to protect against spurious opening of the pressurizer PORVs. Since the inadvertent operation can be mitigated by a Control Room action, the failure is not associated with a VFDR and the modification is only for risk reduction purposes.

The Fire Risk Summary for Area A36 (pg C-376) is revised to remove reference to "proposed cable protection in AttachmentS."

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SSARAI 02 LAR Attachment C, for Fire Area A31, contains generic VFDR resolutions that state, in part, "This VFDR is associated with main control room abandonment actions as listed in Fire Risk Evaluation ... "With the exception of VFDRs A31-08, -11, -22, and -25, each of the VFDRs associated with this fire area have a corresponding recovery action in LAR Attachment G, Table G-1. The licensee stated that recovery action review results are documented in R2167-1012-02, "Recovery Actions Transition Report." Upon further review, the staff noted that the licensee's analysis indicates that recovery actions A31-11 and A31-22 are required.

Explain why the actions for VFDR A31-11 and A31-22 are not included in LAR Attachment G, Table G-1.

NextEra Response The VFDRs associated with 125VDC Panels D-21 (A31-11) and D-22 (A31-22) do not correspond directly with Recovery Actions, but are represented in the Recovery Actions list {Table G-1 of LAR, Reference 1).

Panels D-21 and -22 support multiple devices in the Nuclear Safety Capability Assessment (NSCA) model. On loss of power, most of the supported loads fail in such a fashion so as not to challenge the shutdown strategy credited in the Control Room (i.e., main control room abandonment). Two loads, however, fail in undesirable states: the MSIVs and "SI-RESET" (dummy component to represent the ability to Block and Reset a Safety Injection signal, whether fire-induced or plant-transient-related).

Main Steam Isolation Valves (MSIVs) (discussion for 1MS-02017I 2018; U2 MSIVs are similar but require 125VDC panels D-18 and D-22)

There are 2 solenoid valves in series which supply air to the valve actuator and 2 solenoid valves in parallel which vent air from the valve actuator. The supply solenoid valves are normally open and the vent solenoids are normally closed. All solenoid valves reposition when energized. D-16 energizes one of the supply solenoids (a) and one of the vent solenoids (c). D-21 energizes the other supply and vent solenoid valves (b and d). Therefore, either power supply (D-16 or D-21) will close the main steam isolation valve, since only 1 supply and 1 vent solenoid are required to close the valve. Both D-16 and D-21 (D-18 and D-22) panels are located in the Control Room and are unavailable to effect MSIV closure, for fires that occur in the Control Room.

Local actions to isolate and vent instrument air to failed-close the MSIVs are included in Table G-1 (pages G-11 [U1] and G-19 [U2]); VFDRs A31-04 and A31-18, respectively.

51-RESET "SI-RESET" is a dummy component to (a) evaluate the undesirable impact of a Safety Injection signal using support equipment logics in the NSCA model, and (b) evaluate the ability to Block and Reset a Safety Injection signal. Supporting power is required for SI-RESET IDs as follows:

1-SI-RESET-A D-16 1-SI-RESET-B D-21 2-SI-RESET -A D-22 2-SI-RESET-B D-18 As noted above, panels D-16, D-18, D-21 and D-22 are all located in the Control Room. The undesirable consequences of the SI-RESET device failures are:

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Inability to block/ reset a spurious start of Containment Spray Pumps; local action to trip the CS pumps are included in Table G-1 (pages G-14 [U1] and G-21 [U2]); VFDRs A31-05 and A31-19, respectively.

Inability to block/ reset a load shed of Battery Chargers; local action to position emergency close switch 43/24212B in 'CLOSE' is included in Table G-1 (pages G-24); VFDR A31-10.

Although the VFDRs identifying the loss of 125VDC panels D-21 and D-22 are not directly associated with Recovery Actions in the LAR, the undesirable impacts of their loss are addressed by Recovery Actions.

SSARAI 03 For those fire areas that credit electrical raceway fire barrier system (ERFBS) as described in LAR Attachment C:

a) Identify the VFDRs, if any, that credit the ERFBS for disposition.

b) If credited for dispositioning a VFDR, provide a discussion of the analysis or basis for the acceptability of the ERFBS in resolving the VFDR.

NextEra Response ACRONYMS:

ERFBS Electrical Raceway Fire Barrier System SLERD Refers to the Required Fire Protection Systems & Features table in Attachment C (B-3 Table), so called based on column headings SSAR Safe Shutdown Analysis Report VFDR Variance From Deterministic Requirements a) Currently-installed ERFBS cannot be associated with a VFDR.

Use of the term "ERFBS" in LAR Attachment C (Reference 1, B-3 Table) reflects its usage in the context of the safe shutdown analysis. Where ERFBS is credited (based on SSAR Table 5-7 and confirmed by the safe shutdown analysis), the Required Fire Protection Systems and Features (a.k.a. SLERD Table) indicates Yes (Y) in the Separation (S) column, and the ERFBS is considered an existing fire protection feature, and as such is not associated with a VFDR. Where the ERFBS is not credited (SSAR Table 5-7) or not required for safe shutdown analysis compliance in the area, the SLERD Table indicates No (N) in the Separation (S) column. Therefore, "ERFBS" in the safe shutdown sense (either existing fire protection feature or not credited) is not associated with VFDRs.

A VFDR that is dispositioned by planned cable protection (e.g. ERFBS) would be considered a MOD and reflected in LAR Table S-2, see RAI SSA-03b.

b) The following VFDRs are dispositioned by proposed modifications to protect cable(s); these modifications may utilize raceway protection (ERFBS) or explore other cable protection options (e.g. fire rated cable):

• A01-B-64 in LAR Attachment C (pg. C-38), refer to MOD-20 in LAR Attachment S (pg. S-11)

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• A15-16 in LAR Attachment C (pg. C-210), refer to MOD-11 in AttachmentS (pg. S-7)

• A30-06 in LAR Attachment C (pg. C-324), refer to MOD-16 in AttachmentS (pg. S-9)

• A30-22 in LAR Attachment C (pg. C-324), refer to MOD-16 in AttachmentS (pg. S-9)

Modification design processes (i.e., fire protection and safe shutdown design reviews) will ensure compliance with NFPA 805 Section 4.2.3 for acceptability of resolving each VFDR.

Fire Protection Engineering (FPEl RAI 01 NFPA 805 Section 3.4. 1 (c) requires that the fire brigade leader and at least two brigade members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. LAR Attachment A states that fire brigade members are plant operators and "qualifications of individuals in the fire protection organization are administratively controlled to ensure qualification of the individual commensurate with the position being held and activities being performed." Based on the staffs review, the LAR does not provide sufficient information for the staff to determine whether the brigade leader qualifications demonstrate the competence to assess the potential safety consequences of a fire and advise control room personnel.

Provide additional description of how the fire brigade leader and members have sufficient training and knowledge of nuclear safety systems and understand the effects of fire and fire suppression on nuclear safety performance criteria.

NextEra Response A site fire brigade of at least five members is maintained on site at all times. The plant fire brigade consists of plant operations personnel independent of the shift manager and 3 members of the minimum shift crew necessary for the safe shutdown of the plant. The fire brigade leader is typically an Operations Supervisor, but this can be delegated to a Lead Auxiliary Operator or his designee. Auxiliary Operator minimum training is specified below.

Auxiliary Operators (AO's) are required to be qualified under the Auxiliary Operator Program. The purpose of this program is to establish the requirements for initial and continued training. Initial training ensures entry-level personnel attain the required knowledge and skills to perform the duties of an Auxiliary Operator. Continued training ensures that incumbents maintain and improve performance. One of the responsibilities of being an Auxiliary Operator is to be a qualified fire brigade member.

Fire Brigade Member qualifications include the completion of training and evaluation of the subject matter listed below.

Classroom training courses I Lessons completed for Auxiliary Operators include the following:

Primary System Secondary Systems Electrical Systems Administrative Training Health Physics Fuel Handling Page 47 of 58

Plant Systems Plant Operating Procedures Teamwork and Human Performance Tools Required reading includes:

Industrial Fire Brigade Training Manual NP 1.9.9, Transient Combustible Control NP 1.9.13, Ignition Control Procedure NP 1.9 .14, Fire Protection Organization NP 8.4.11, Penetrating Fire Barriers Fire Attack Plans (FAPs)

Brigade Training FOP 1.1 Potential Fire Affected Safe Shutdown Components FOP 1.2 Familiarization with the operation and location of fire protection equipment.

Familiarization with Fire Protection System surveillance tests.

The Auxiliary Operator Training Program provides the fire brigade leader and members with the training and knowledge of nuclear safety systems and potential interactions with fire and fire suppression systems on nuclear safety performance criteria.

FPE RAJ 02 NFPA 805 Section 3.3.5.2 requires that only metal trays and metal conduits be used for electrical raceways. In LAR Attachment A, the compliance basis states that in limited circumstances, exposed conduits have thin plastic coating and clarifies that these plastic coated metal conduits are considered as

' complying with this section of NFPA 805 because the base material is metal. The staff notes that metal trays and metal conduit do not contribute to fire propagation.

a) Provide additional justification that demonstrates this conduit is equivalent to metal conduit in fire propagation behavior or request approval using a performance-based method in accordance with 10 CFR 50.48(c)(2)(vii).

b) Discuss the extent of condition of the use of thin plastic coated conduits (i.e., further describe the "limited circumstances" of use).

c) Describe whether these conduits present an exposure hazard to other safe shutdown circuits or if the conduit could propagate fire to locations containing safe shutdown circuits.

d) Describe whether the fire propagation/exposure potential of these conduits are considered in the assessment of fire damage.

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NextEra Response a) The use of plastic-coated metallic conduit in the plant is limited to reactor nuclear instrumentation circuits. The plastic coating prevents grounding of the conduit in order to prevent "noise" in the circuitry. This conduit is coated with an exceptionally thin layer of plastic material, the amount of which is negligible such that it is not expected to sustain fire on the conduit exterior for any credible length of time.

As identified in the compliance basis for NFPA 805 Section 3.3.5.2 in LAR Attachment A (Reference 1), the conduits in question are metallic, which meets the code requirement. The plastic coating utilized on these conduits is very thin, estimated 0.060", and is not expected to provide any credible influence on fire propagation behavior.

b) A review of the PBNP cable and raceway database revealed 26 metal conduits with thin plastic coatings. All are associated with the reactor nuclear instrumentation. Per site drawing notes, the plastic coating is to prevent grounding of the conduit. Nuclear instrumentation subject matter experts have indicated that this helps prevent "noise" in the circuitry. The 26 conduits are listed below:

1C208-8 1S604 2PVC-B 2S606 1PVC-B 1S606 2PVC-R 2S608 1PVC-R 1S608 2PVC-W 2S609 1PVC-W 1S610 2PVC-Y 2S610 1PVC-Y 1S612 2S600 2S611 1S600 1S614 2S602 1S602 2C208-7 2S604 Conduit 1C208-8 is a 2" diameter conduit. It is routed from a containment penetration on the 8' elevation, through Pipeway #1, through the Primary Auxiliary Building (PAB), and terminates at a panel in the south Aux Feed Pump Room.

Conduits 1PVC-B, R, W, and Yare 3" diameter conduits. They are routed from containment penetrations on the 8' elevation, through one of the 8' elevation Pipeways (R & B Pipeway #1 and W & Y Pipeway #2), through the PAB, into the Cable Spreading Room (CSR) in the southwest corner, along the CSR south wall, and terminate at junction boxes on the CSR ceiling south end.

The 1S series conduits are 3" diameter conduits. They are all located in the Unit 1 Containment, and are routed from junction boxes in containment to junction boxes near containment penetrations.

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The corresponding Unit 2 conduits in the list above are of the same diameter and follow similar Unit 2-side routes (i.e., Conduit 2C208 terminates at a panel in the north Aux Feed Pump Room, and Conduits 1PVC-8, R, W, andY enter the northwest corner of the CSR, travel along the north CSR wall, and terminate at junction boxes on the CSR ceiling north end).

c) As identified in the response to part (a), the properties of this plastic-coated conduit do not lend themselves to credible fire propagation that would cause safe shutdown circuits or equipment to be affected. If a fire were to occur in a fire area containing these conduits, existing controls such as fire-rated barriers, electrical raceway fire barrier systems, spatial separation, etc. would ensure that redundant cabling and circuitry would not be affected by the fire.

d) The coating of plastic on select conduits in these areas would not impact the ability of the plant to safely shutdown. Safe shutdown circuits that are within these plastic-coated conduits are appropriately routed in metallic conduit, and would not have any interaction with the plastic coating. The fire propagation/exposure potential of these coated conduits is not considered in the assessment of fire damage.

FPE RAI 03 In LAR Attachment A, the compliance statement for elements 3.6.1, 3.9.1 (2), and 3.1 0.1 (2) is "Complies with Clarification." However, in the "Compliance Basis," reference is made to documentation that provides justification for deviations from the NFPA codes associated with these elements, which appears to be an engineering evaluation per NFPA 805.

Provide an explanation why the compliance statement is not characterized as "Complies with use of EEEEs?"

NextEra Response LAR Section 4.1.1 (Reference 1), "Overview of Evaluation Process," describes the review of each section and subsection of NFPA 805 Chapter 3 against the PBNP Fire Protection Program. Section 4.1.1 identifies the various compliance statements. Elements cited in this RAI where the compliance statement indicates "Complies with Clarification" are related to NFPA Code Conformance Assessments. During the development of the B-1 table, it was decided to consider the Code Review compliance as "Compliance with Clarification". The code assessments conclude PBNP complies with the particular NFPA code, however, portions that deviate from the code were determined to be acceptable based on the results of a Fire Protection Technical Evaluation. Since conformance is based on an engineering evaluation, the Table B-1 compliance statement for Elements 3.6.1, 3.9.1 (2) and 3.1 0.1 (2) will be changed from "Compliance with Clarification" to "Complies with Use of EEEEs."

LAR, Attachment A, Pages A-84, A-98 and A-1 04, is revised to change the compliance statement to "Complies with Use of EEEEs." A markup of applicable LAR pages is provided in Attachment D.

FPE RAJ 04 The exception to NFPA 805 Section 3.11.4(b) requires conduit with inside openings of 4-inches or less in diameter to be sealed at the fire barrier unless the conduit extends greater than 5-feet on each side of the wall, in which case the opening must be provided with a smoke and hot gas seal. In LAR Attachment A, Page 50 of 58

for this element, clarification is provided that states that small conduits provided for items such as lighting circuits, are embedded in the concrete construction, are not considered paths for the spread of fire, and have not been sealed. NFPA 805 Section 3. 11.4(b) does not specify an exception for small conduit.

Based on the NRC staffs review, the LAR does not provide sufficient information for the staff to determine whether the described configuration is acceptable.

Describe how compliance with NFPA 805 Section 3.11.4(b) will be achieved for these small conduits.

NextEra Response In response to the differing conduit fire seal criteria guidance in BTP 9.5-1 Appendix A, NUREG 0800 and 10 CFR 50.48 Appendix R and in the interest of providing cost-effective fire protection based on sound technical information, a group of 23 nuclear utilities sponsored by Wisconsin Electric (WE) embarked on a research project back in 1986 to evaluate internal conduit sealing requirements and develop design guidelines based on fire test data.

Results of this project were documented in a proprietary report entitled "Conduit Fire Protection Research Program." This report, prepared by Professional Loss Control, Inc, was submitted to the NRC on June 19, 1987 for review. On May 18, 1988 the NRC requested additional information (RAI) to complete their review. The WE response to the RAis was provided back to the NRC on June 15, 1988.

On October 23, 1989, the NRC issued the results of their review of Conduit Fire Protection Research Program in a Technical Evaluation Report (TER) and accompanying Safety Evaluation. The TER was prepared by technical assistance contractor Science Applications International Corporation (SAl C).

Based on the staffs review, it was concluded that the subject research program justifies the .

implementation of revised conduit fire seal criteria.

LAR Attachment A, Section 3.11.4(b) Compliance Basis column quotes conduit fire seal criteria from Fire Protection Report, Section 5.1.3.4. The FPER quote is consistent with Conduit Fire Protection Research Program results and FPTE 001.

LAR Attachment A, Section 3.11.4(b) "Compliance Statement" will be revised from "Complies with Clarification" to "Complies by Previous NRC Approval," which more accurately reflect how compliance with NFPA 805 Section 3.11.4(b) is achieved for small conduits. The "Compliance Basis" column will be revised to provide additional information related to the development and approval of the PBNP small conduit fire seal criteria. The Reference Document column is revised to add supporting documentation associated with the development of these criteria. See attached LAR Attachment A Section 3.11.4(b) markups and inserts in Attachment E.

FPE RAI 05 LAR Section 4.1.2.3 and LAR Attachment A identify that NRC approval is requested per 10 CFR 50.48(c)(2)(vii) for a deviation to NFPA 805 Section 3.5.6. LAR Attachment A indicates this request is addressed in LAR Attachment L, Approval Request 6. Approval Request 6 appears to address the short circuit rating of the circuit breaker as required by NFPA 20. With regard to start and stop controls for fire pumps, Approval Request 6 only contains the citation of the NFPA 805 requirement but does not provide the basis for approval, or the acceptance criteria evaluation for this element of NFPA 805.

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To address the remote start and stop capability of the fire pump, provide the necessary basis for approval and evaluation of the impact on nuclear safety performance criteria, radioactive release criteria, safety margin and defense-in-depth as required by 50.48(c)(2)(vii).

NextEra Response The electric-motor-driven and diesel-engine-driven fire pumps meet the requirements of NFPA 805 Section 3.5.6 by design. They are both configured to start automatically on low system pressure and can also be started manually; and they must be shut down manually at their respective local control panels.

Clarification of the applicability of NFPA 805 Section 3.5.6 to this request for NRC approval is as follows:

The electric-motor-driven and diesel-engine-driven fire pumps are both provided with local manual stop only, by design. However, there are two instances by which the electric-motor-driven fire pump can stop automatically, without manipulation of the local fire pump controls:

An inadvertent automatic stop of the electric-motor-driven fire pump could occur upon a trip of the existing class IE emergency power supply circuit breaker, the potential of which is the subject of this request for NRC approval. As discussed in the original transition report, on Page L-23, "In the unlikely event of a circuit breaker trip, administrative procedures require Operations to immediately investigate the condition and restore this power supply when it is confirmed it is electrically safe to perform such action. In the interim, a diesel-driven fire pump is available and is designed to automatically start if the electric pump has stopped." In summary, a circuit breaker trip would cause an automatic stop of the electric fire pump, contrary to the intent of NFPA 805 Section 3.5.6; and, once the circuit breaker is tripped, it needs to be reset manually before the pump will automatically restart, also contrary to the intent of NFPA 805 Section 3.5.6.

A safety injection signal for the Engineered Safety Features (ESF) actuation system would automatically trip the electric fire pump as part of an automatic emergency diesel generator (EDG) load shed sequence to prevent high starting currents from overloading the EDG upon startup. The basis for the acceptability of this condition is equivalent to that for an automatic stop due to a trip of the power supply circuit breaker, as discussed in the request for NRC approval in the original transition report. As stated above,

" ... a diesel-driven fire pump is available and is designed to automatically start if the electric fire pump has stopped."

Although the electric fire pump complies with NFPA 805 Section 3.5.6 by design, the potential circuit breaker trip and safety injection signal scenarios do cause the pump to deviate from the code requirement. The justification for these conditions, including the basis for NRC approval and the evaluation of the conditions' impacts on nuclear safety performance criteria, radioactive release criteria, safety margin, and defense-in-depth, is provided in the original transition report.

FPE RAI 06 NFPA 805 Section 3.3.5.3 requires that electric cable construction comply with a flame propagation test acceptable to the AHJ (i.e., NRC). LAR Attachment A provides the basis for previous NRC approval, which includes a statement (from the August 2, 1979 NRC SER) that a qualified flame retardant coating or material will be used wherever there is a concern for rapid flame propagation from an electrical fire that could compromise redundant safety-related divisions. This element in LAR Attachment A also identifies Implementation Item, IMP-136, from LAR AttachmentS, Table S-3, to incorporate this original SER requirement into plant design guidelines.

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Since there is reliance on the original1979 SER approval as the basis for compliance with this element, explain how existing configuration control mechanisms have maintained compliance to the original SER.

NextEra Response The August 2, 1979 NRC SER approved the use of installed cable that was not qualified per the IEEE 383-1974 edition. The approval acknowledged that wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions, a qualified flame retardant coating or material would be used to minimize the likelihood.

Section 3.0 of Point Beach Fire Protection Technical Evaluation FPTE 011, "Technical Evaluation of Acceptable PBNP Cable Fire Retardancy Standards Power, Control and Specialty Cable," states, in part, "The cable originally installed in PBNP was not qualified per the fire retardancy standard IEEE 383-1974.

New cable installations are qualified per fire retardancy standards IEEE 383, UL-91 0, UL-1581, NFPA-262, or equivalent cable fire retardancy standards .... " This document and a summary of its contents are also provided in the Fire Protection Evaluation Report (FPER). With the guidance of FPTE 011 in place in the fire protection program for use when performing plant changes involving cable installation, compliance with the 1979 commitment to use a qualified flame retardant material is maintained. The intent of IMP-136 is to ensure that Design Guideline DG-E13, "Insulated Electrical Cable Installation,"

adequately reflects the NRC commitment. IMP-136 will be reworded as follows, to better clarify the intent:

IMP-136: Design Guideline DG-E13, "Insulated Electrical Cable Installation," is revised to make reference to FPTE 011, "Technical Evaluation of Acceptable PBNP Cable Fire Retardancy Standards Power, Control and Specialty Cable," which provides the acceptable cable qualification standards to meet the August 2, 1979 NRC commitment to use a qualified flame retardant coating or material wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions.

A markup of the LAR Attachment A and AttachmentS, Table S-3 is provided in Attachment F.

FPE RAI 07 LAR Attachment L, Approval Request 5, addresses deviations from NFPA 805 Section 3.3. 7.1 with regard to code required separation between the hydrogen storage and the adjacent turbine building (TB). Provide the following:

a) A set of diagrams depicting the layout and cross sections of the hydrogen storage facility and the TB wall with sufficient detail to understand the separation distances, openings and any protective features being credited.

b) An explanation of the apparent discrepancy between the 20. 7 feet and 35 feet distances from the hydrogen storage system and the TB east wall. Under the heading, "Hydrogen System Configuration,"

the horizontal distance between the hydrogen storage system and the east TB wall is cited as 20. 7 feet.

Under the heading, "Code Requirements Summary," second bullet, the last sentence states, 'The Turbine Building east wall and openings located at grade are approximately 35 feet away from the hydrogen system."

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c) A description of how the openings above the hydrogen storage system meet the separation requirements. The second bullet under "Code Requirements Summary" states, "Table 2 a/so specifies that the Turbine Building wall openings, such as the doors and louver located at grade, be separated from the hydrogen system by a minimum 10 feet when the openings are not above any part of the hydrogen system." The description of the hydrogen system configuration identifies louvers below the roof-line and roll-up doors in east wa/1 of the TB "above the hydrogen storage location".

NextEra Response a} Diagrams and photos depicting the Hydrogen Storage Tank Area, the east wall of the Turbine Building and the opening in this wall are provided in Attachment G. The distance measured horizontally from the closest Hydrogen Storage Tank to the East wall of the Turbine Building was confirmed to be 20.7' which is less than 25' required by NFPA 567, Standard for Gaseous Hydrogen Systems at Consumer Sites. The distance from the Hydrogen Storage Tanks to the closest opening on the East Wall not above any part of the system (which is a 3'- 8" louver located south of the H2 Storage Tanks and adjacent to the Rollup Door) was confirmed to be 35',

which is greater than 10' required by NFPA 567, Standard for Gaseous Hydrogen Systems at Consumer Sites. There are no openings on the east wall of the Turbine Building located directly above the Hydrogen Tanks. The closest opening on the east wall of the Turbine Building that is above the elevation of the upper Hydrogen Tanks is a roll up door which is located approximately 13' north and 12' above of the H2 Storage Tanks. Measurements from the upper west H2 tank indicated a distance of 24' to the east wall of the Turbine Building directly below the rollup door and 12' vertically up to the rollup door. The line-of-sight distance between this rollup door and the closest Hydrogen tank is calculated to be 26.8'. This distance is greater than the 25' required by NFPA, if the opening were directly above the system. There are no protected features being credited in LAR Attachment L (Reference 1), Approval Request 5.

b) Diagrams and photos depicting the Hydrogen Storage Tank Area, the east wall of the Turbine Building and the opening in this wall provided in Attachment G. The distance measured horizontally from the closest Hydrogen Storage Tank to the East wall of the Turbine Building was confirmed to be 20.7'. The distance from the Hydrogen Storage Tanks to the closest opening on the East Wall not above any part of the system (which is a louver located south of the H2 Storage Tanks and adjacent to the Rollup Door) was confirmed to be 35'.

The statement in LAR Attachment L for Approval Request 5, Code Requirements Summary, second bullet, should have read, 'The Turbine Building east wall aRd openings located at grade are approximately 35 feet away from the hydrogen system."

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c) NFPA 567, Table 2 requires wall openings "Above any part of a system" to be separated by 25' for Hydrogen Systems in excess of 15,000 CF ." There are no openings on the east wall of the Turbine Building located directly above the Hydrogen Tanks. The closest opening on the east wall of the Turbine Building that is above the elevation of the upper Hydrogen Tanks is a roll up door which is located approximately 13' north and 12' above of the closest Hz Storage Tank.

Measurements from the upper west Hz tank indicated a distance of 24' to the east wall of the Turbine Building directly below the rollup door and 12' vertically up to the rollup door. The line-of-sight distance between this roll up door and the closest Hydrogen tank is calculated to be 26.8'.

This distance is greater than the 25' required by NFPA, if the opening were directly above the system. All other openings on the east wall are not directly above the Hydrogen system and are separated from the Hydrogen System in excess of 25'.

FPE RAI 08 LAR Attachment A, Section 3. 11 .3, describes the compliance basis for fire barrier penetrations and identifies "Compliance by Previous Approval" for installed water curtains that protect doorless entrances through fire separations in various locations. Water curtains are also described in LAR Table 4-3 as required fire protection features in several areas. The approval basis is cited as exemption request dated July 3, 1985. LAR Attachment K does not indicate that this licensing action (i.e., exemption request) is being transitioned in support of NFPA 805 compliance.

Describe whether this approval basis is being transitioned in support of NFPA 805 and the reasoning for the decision.

NextEra Response LAR, Attachment A (Reference 1), NEI 04-02 Table B-1, Transition of Fundamental Fire Protection Program & Design Elements, and Section 3.11.3, Fire Barrier Penetration, indicates PBNP complies with clarifications. The Compliance Basis indicates water curtains were provided to protect entranceways in fire barriers that did not have rated fire doors. The areas identified as having the water curtains are the Unit 1 Motor Control Center Room, Component Cooling Water Pump Room, Unit 2 Motor Control Center Room and Safety Injection and Containment Spray Pump Room. Water curtains were installed to support exemptions from the requirements of 10 CFR 50 Appendix R III.G.2.b in these areas.

LAR Section 4.2.3 and Attachment K, Existing Licensing Action Transition, identifies the exemptions in these areas that credit the installed water curtains.

Licensing Action Number 2 is for the exemption in the Unit 1 Motor Control Center Room (Fire Zone 1, which has been revised to Fire Zone 156 in Fire Area A06). This exemption is no longer required for transition since the fire area is transitioning with NFPA 805, Section 4.2.4.2, Performance-Based Approach- Fire Risk Evaluation. Separation and lack of full area automatic suppression are addressed through Fire Risk Evaluations.

Licensing Action Number 3 is for the exemption in the Component Cooling Water Pump Room (Fire Area A01-A). This exemption is no longer required for transition, since the fire area is transitioning with NFPA 805, Section 4.2.4.2, Performance-Based Approach- Fire Risk Evaluation. Separation and lack of full area automatic suppression are addressed through Fire Risk Evaluations.

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Licensing Action Number 4 is for the exemption in the Unit 2 Motor Control Center Room (Fire Zone 4, which has been revised to Fire Zone 166 in Fire Area A 15). This exemption is no longer required for transition, since the fire area is transitioning with NFPA 805, Section 4.2.4.2, Performance-Based Approach- Fire Risk Evaluation. Separation and lack of full area automatic suppression are addressed through Fire Risk Evaluations.

Licensing Action Number 6 is for the exemption in the Safety Injection and Containment Spray Pump Room (Fire Zone 2, which has been revised to Fire Zone 151 in Fire Area A02). This exemption is no longer required for transition because an automatic sprinkler system has been provided for complete area coverage and the fire area is transitioning under NFPA 805, Section 4.2.4.2, Performance-Based Approach - Fire Risk Evaluation. Separation is addressed through Fire Risk Evaluations.

Since the exemptions are not being transitioned and the water curtains are not being credited for separation in these areas, Table 4-3 indicates "N" under "Required?"; "S" column.

The water curtains in these fire areas are credited for engineering equivalency evaluation and fire risk.

Table 4-3 indicates "Y" under "Required?"; "E" and "R" columns for these fire areas.

Therefore, although the exemptions are no longer necessary to support the deterministic separation criteria, the "Compliance by Previous Approval" is requested in LAR Attachment A, Section 3.11.3, to support reliance on the water curtains in the engineering equivalency evaluations and for fire risk.

Programmatic (PROG) RAI 01 Based on the NRC staff's review of the LAR and associated documentation it was determined that the LAR did not provide the information needed for the NRC staff to evaluate what changes will be made to the site QA program to incorporate NFPA 805 requirements.

a) Describe the changes that are anticipated that would modify the PBNP QA program to ensure NFPA 805 fire protection requirements are incorporated into existing processes and programs.

b) Discuss how NFPA 805 Section 2. 7.3 requirements are included within and implemented by the existing PBNP QA program and any planned modifications.

NextEra Response a) The Point Beach Quality Assurance Manual is maintained as part of the NextEra fleet process. The requirements are contained in the fleet Quality Assurance Topical Report, FPL-1. In response to Duane Arnold Energy Center NFPA 805 implementation activities, this report was updated to revision 14 and released on November 21, 2013. Specific sections revised included Section A.7, Regulatory Commitments for plants with an NFPA 805 Fire Protection licensing bases, which states: "For plants with an NFPA 805 fire protection licensing bases, NextEra Energy commits to implement Regulatory Guide 1.205, December 2009, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," which endorses in part, NEI 04-02, Revision 2, Nuclear Energy Institute Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c). The implementation of these documents is described in the station specific Technical Specifications and License Conditions and station specific NRC approved Safety Evaluation Reports."

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NextEra Fleet procedures already adopted by Point Beach that have been updated to include NFPA 805 requirements include: FP-AA-104, Fire Protection Program; EN-AA-105-1000, PRA Configuration Control and Model Maintenance; and EN-AA-202-1 001, Engineering Change Scoping and Screening.

Any changes to the scope of required fire protection systems and features as a result of NFPA 805 will be addressed as part of creating the monitoring program as specified in IMP- 139 of Table S-3 of the NFPA 805 LAR (Reference 1).

As discussed in the response to part a) of this RAI the Point Beach Quality Assurance Manual is maintained as part of the Next Era fleet process and has been updated as part of the Duane Arnold Energy Center NFPA 805 implementation activities.

Since the fleet QA program was modified to address NFPA 805 requirements as part of the Duane Arnold Energy Center NFPA 805 implementation activities, there are no planned modifications to the program specifically to address Point Beach.

b) As discussed in the response to part a) of this questions the Point Beach Quality Assurance Manual is maintained as part of the NextEra fleet process and has been updated as part of the Duane Arnold Energy Center NFPA 805 implementation activities.

Since the fleet QA program was modified to address NFPA 805 requirements as part of the Duane Arnold Energy Center NFPA 805 implementation activities, there are no planned modifications to the program specifically to address Point Beach.

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References:

(1) NextEra Energy Point Beach, LLC, letter to NRC, dated June 26, 2013, "License Amendment Request 271, Transition to 10 CFR 50.48(c)- NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition" (ML131820453)

(2) NRC e-mail to NextEra Energy Point Beach, LLC, dated September 9, 2013, "Request for Supplemental Information Regarding the Acceptability of the Proposed Amendment Request" (ML13256A197)

(3) NextEra Energy Point Beach, LLC, letter to NRC, dated September 16, 2013, "License Amendment Request 271 Supplement 1 Transition to 10 CFR 50.48(c)- NFPA 805" (ML13259A273)

(4) NRC letter to NextEra Energy Point Beach, LLC, dated September 25,2013, "Point Beach, Units 1 and 2 -Acceptance Review of Licensing Action re: License Amendment Request to Transition to NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (ML13267A037)

(5) NRC e-mail to NextEra Energy Point Beach, LLC, dated July 8, 2014, "Point Beach Nuclear Plant, Units 1and 2- Final (Revised) Requests for Additional Information re: License Amendment Request Associated with NFPA 805 (TAC Nos. MF2372 and MF2373)" (ML14189A365)

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ATTACHMENT A LAR UPDATES FOR PBNP RAI PRA 14

NextEra PBNP Attachment H - NEI 04-02 FAQs Summary Table This table includes the approved FAQs that have not been incorporated into the current endorsed revision of NEI 04-02 and utilized in this submittal:

Table H NEI 04-02 FAQs Utilized in LAR Submittal No. Rev. Title FAQRef. Closure Memo 06-0008 9 NFPA 805 Fire Protection ML090560170 ML073380976 Engineering Evaluations 06-0022 3 Acceptable Electrical Cable ML090830220 ML091240278 Construction Tests 07-0030 5 Establishing Recovery Actions ML103090602 ML110070485 07-0032 2 Clarification of 10 CFR 50.48(c), ML081300697 ML081400292 10 CFR 50.48(a) and GDC 3 clarification 07-0035 2 Bus Duct Counting Guidance for ML091610189 ML091620572 High Energy Arcing Faults 07-0038 3 Lessons learned on Multiple ML103090608 ML110140242 Spurious Operations 07-0039 2 Lessons Learned- NEI B-2 ML091420138 ML091320068 Table 07-0040 4 Non-Power Operations ML082070249 ML082200528 Clarification 08-0042 0 Fire Propagation from Electrical ML080230438 ML092110537 Cabinets ML091460350 08-0043 1 Electrical Cabinet Fire Location ML083540152 ML092120448 ML091470266 08-0044 0 Large Oil Fires ML081200099 ML092110516 ML091540179 08 0046 Q lnoipient Fire Detection Systems MbQ8~~QO~~O Mb093~~0426 Mb093220~97 08-0047 1 Spurious Operation Probability ML082770662 ML082950750 08-0048 0 Fire Ignition Frequency ML081200291 ML092190457 ML092180383 08-0049 0 Cable Fires ML081200309 ML092100274 ML091470242 08-0050 0 Non Suppression Probability ML081200318 ML092190555 ML092510044

ATTACHMENT 8 LAR UPDATES FOR PBNP RAI PRA 21

Attachment B- PRA RAI 21 IMP-142 1,2 The Fire PRA model will be updated after all modifications and implementation items are 4.8.2 complete and as-built. If the revised Fire PRA shows a risk increase of greater than 1E-07 for CDF or 1E-08 for LERF, then the results will be entered into the corrective action program to determine the cause of the risk increase and determine corrective actions.

This is being tracked by NAMS Action Request 1882226.

ATTACHMENT C LAR UPDATES FOR PBNP RAI SSA 01 c

NextEra PBNP Attachment B- NEI 04-02 Table B-2 -Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location required to meet Appendix R requirements. The issues were reviewed for compliance to NFPA 805 and no additional changes were identified. Refer to Action Request Nos. AR # 01298593, AR # 01298594 for disposition of those instances where there was a lack of coordination. See Alignment Basis for 3.5.2.5 for a discussion of loss of breaker coordination due to loss of DC control power.

480V Switchgear Calculation 2001-0049 identified nine instances where there was a lack of coordination. New breaker settings 1111ere identified by the Calculations Reconstitution Project and CAP #028771 (NAMS AR 01224052) was initiated. This CAP is shown as being complete and closed 3/11/2003. With the recommended breaker settings made, the breaker coordination for the 480 volt switchgear busses evaluated in this calculation is complete. In addition, Action Requests 01224052 and 01339558 were issued to disposition many of the issues identified in the calculation. The dispositions were reviewed and there are no additional impacts on NFPA-805.

480V Motor Control Centers Calculation 2004-0030 identified various instances where coordination could not be demonstrated. In each of these cases, the technical evaluation review was able to identify a technical justification to support that there is no impact to NFPA-805 compliance.

Full selective coordination does not exist between the supply breaker to MCC B-21 and all load breakers. Modify MCC B-21 supply breaker settings to provide full coordination. Refer to AttachmentS Table S-2 MOD-26-1.

120VAC Distribution Panels (Addressed in FPTE-2007 -001)

The technical evaluation review identified a potential coordination issue associated with 120VAC distribution panels 1Y103, 1Y104, 2Y103 and 2Y104. Each of these panels contain a 100 ampere breaker associated with Radiation Monitoring System. Since these branch breakers are large thermal magnetic breakers they require long time durations to trip open on low fault levels.

The inverter output breakers associated with powering these distribution panels are designed to trip when fault levels reach 260 amperes or more for a duration in excess of 5 seconds. Since the inverters are capable of generating fault currents in this order of magnitude, this condition could result in the tripping of the associated inverter output breaker upon a fault of the cable{s) providing power to the Radiation Monitoring System. As a result, AR # 01316511 has been initiated to identify this issue. See AttachmentS Table S-2 MOD-26-3.

PBNP does not have a single calculation which addresses both safe shutdown and non-safe shutdown branch feeder protective devices with respect to coordination. As a result, it was necessary for the technical evaluation review to review the various distribution panels required for safe shutdown. With exception of the four 120VAC buses identified above, the technical Revision 0 Page B-85

NextEra PBN P Attachment B- NEI 04-02 Table B-2 -Nuclear Sa1fety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location evaluation was successful in demonstrating that most of the circuits from these panels are routed in alternate shutdown areas or that electrical coordination exists. Since many of the cables are in alternate shutdown fire areas, the failure of these circuits is not of concern with respect to the failure of these power supplies. PBNP has alternate shutdown capability which is independent of these fire areas.

Full selective coordination does not exist regarding the 120VAC safety related instrument busses, as documented in Sargeant and Lundy Report, "Safe Shutdown 120VAC Distribution Panel Coordination Evaluation". Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-26-3.

125VDC Distribution Panels A review of calculation N-92-005 Revision 2 concluded that the only lack of coordination potentially impacting NFPA-805 compliance is the lack of coordination of fuse D72-14-02 on panel D-14 with fuse D72-02-06 on panel D-02. The replacement of fuse D72-02-06 is to be accomplished by AR # 01232138.

Numerous fuses associated with 13.8KV system 125VDC control power have ratings below the normal operating voltage and require replacement. Refer to AR 01877063.

120VAC and 125 VDC Distribution Panels Calculation no. V878-15-CA-02 performed a review of both the 120VAC and 125VDC distribution panels. The 125 VDC branch circuits are coordinated and, as a result, have no impact on NFPA-805 compliance. For the 120 VAC branch circuits, the inverter backup power supply has high available fault current which results in a lack of coordination. In order to achieve coordination, modifications will be performed. Refer to AttachmentS Table S-2 MOD-26-3.

Revision 0 Page B-86

NextEra PBNP Attachment B- NEI 04-02 Table B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location NEI 00-01 Ref NEI 00.01 Section 3.0 Guidance 3.5.2.5 Circuit Failures Due to The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to fire Common Enclosure damage to a circuit either whose isolation device fails to isolate the cable fault or protect the faulted cable from Concerns reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.

The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.

Applicability Comments Applicable See "Alignment Basis" for exceptions to 'Align with Intent' grade.

Alignment Statement Alignment Basis Reference Documents Aligns with Intent The primary issue is the ignition of a secondary fire which then results in failure of cables local SSAR Section 3.3.2.1 to that fire. FPTE-2007-001 Calculations:

The PBNP SSAR, Section 3.3.2 identifies requirements for the Common Enclosure concern. 2004-0009 Section 6.0 of FPTE-2007-001 provides a qualitative assessment of the Common Enclosure 2001-0049 concern for Point Beach Units 1 and 2. 2004-0030 2005-0005 Overall, the existing PBNP methodology for analysis of Circuit Failures Due to Common N-92-005 Enclosure Concerns is consistent with NEI 00-0'1, Section 3.5, however existing cable protection V878-15-CA-02 issues for proper circuit protection could lead to loss of electrical distribution equipment affecting R2167-1019-001 Section 8.3.4.1 safe shutdown equipment operability.

FPTE-2007-001 was performed to further resolve these common enclosure issues. Summary results are as follows:

13kV & 4kV Switchgear Calculation 2004-0009 either demonstrated adequate cable protection or a technical justification was provided in the evaluation.

Point Beach reviewed the loss of DC control and the effect of that loss on the ability of breakers to clear a fault. The concern is ignition of the faulted cable in a location other than the affected fire area. Point Beach determined that several breakers at the 4 kV and 13 kV level were susceptible to this failure mode. Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-17 and MOD-30.

Revision 0 Page B-87

NextEra PBNP Attachment B- NEI 04*02 Table B*2 ~ Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location 480V Switchgear Calculation 2001-0049 identified several instances where cable protection was in question.

However, it was demonstrated that these cables were either in the same fire area as their associated power supplies or that the fault current only exceeds the thermal withstand for a short distance from the power supply at which point the fault current decreased to an acceptable level.

480V Motor Control Centers Nine cables were i,jentified in Sargent and Lundy Calculation 2013-01785 that are not fully protected for overloads in the short time region. Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-24.

Numerous cables were identified in Sargent and Lundy Calculation 2013-01785 that are not fully protected for short circuits in the instantaneous region. However, the fault current only exceeds the thermal withstand for a short distance from the power supply at which point the fault current decreases to an acceptable leveL Evaluation of cable protection for cables supplied from non-safe shutdown 480V MCCs and power panels, and 208/120V lighting panels is incomplete (Ref. AR 1877063). Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-26-1 and MOD-26-3.

120VAC Distribution Panels Calculation 2005-0005 addresses the thermal withstand capability of power cables directly connected to the individual panel busses. Although a large portion of the cables were shown to have sufficient thermal withstand capability, there were a number cables which were deficient.

As a result, FPTE-2007-001 addressed many of these deficiencies by either identifying those circuits routed in a single fire area or in Alternate Shutdown fire areas only. Only single circuit failed meet one of these conditions. Further analysis was performed on the cable to demonstrate that its thermal withstand capability would not be exceeded for a fire outside of the fire area of the power supply.

125VDC Distribution Panels No issues have been identified regarding protection of 125VDC system cables.

Revision 0 Page B~88

Security-Related Information -Withhold from Public Disclosure Under 10 CFR 2.390 NextEra PBNP Attachment S- Modifications and Implementation Items Table S-2 Plant Modifications Committed 1 In Comp Risk Informed Item Rank Unit Problem Statement Proposed Modification FPRA Measure Characterization MOD-25 L Cable thermal damage Replace #10 Conductors with #8 y N Ensures conductor protection curve intersects the Conductors and 250°C Short in vertical region of breaker vertical region of the Circuit Temperature insulation. characteristic.

breaker characteristics curve. Cable ZB 1401 DA -

1. 1 0 Conductors with 250,C Short Circuit Temperature insulation protected by a 30A HMCP breaker.

MOD-25 L 2 Cable thermal damage Replace existing Conductors with y N Ensures conductor protection curve intersects the Conductors that are protected by in vertical region of breaker vertical region of the existing overcurrent device. characteristic.

breaker characteristics curve. Cable ZD2401 DA

- #1 0 Conductors with 250°C Short Circuit Temperature insulation protected by a 30A HMCP breaker.

MOD-26-1 L 1,2 Coordination is not Revise breaker 2B52-28C settings. y N Ensures breaker coordination maintained between to support Fire PRA analysis Switchgear Breaker 2B52- assumptions.

28C and MCC Breaker B52-211M.

MOD-26-2 L 1,2 Coordination is not Replace fuses to ensure proper y N Ensures breaker coordination maintained for circuits voltage, short circuit rating and to support Fire PRA analysis protected by three DC coordination. assumptions.

fuses. Eight DC fuses do not have appropriate voltage ratings.

Revision 0 Page S-16

Security-Related Information - Withhold from Public Disclosure Under 10 CFR 2.390 NextEra PBNP Attachment S- Modifications and Implementation Items Table S-2 Plant Modifications Committed 1 In Comp Risk Informed Item Rank Unit Problem Statement Proposed Modification FPRA Measure Characterization MOD-26-3 L 1,2 The Inverter backup A modification will be installed to y N Ensures breaker coordination power supply has such resolve the condition that prevents to support Fire PRA analysis high available fault current the Inverter backup power supply assumptions.

that main breaker and from affecting proper breaker load breakers in the 120V coordination. This modification will AC distribution panels do address the condition where Safe not coordinate. Shutdown and Non-Safe Shutdown loads are powered from the same Non-Safe Shutdown and 120VAC distribution panel.

Safe Shutdown loads are powered from the same 120V AC distribution panel breakers that result in Associated Circuit issues in certain Fire Zones.

MOD-27 L 1,2 The fire area boundary is Unsealed penetrations in the floor y y Risk is reduced by crediting the inadequate due to lack of of the non-vital switchgear room to sealed penetration in the fire fire rated penetration the vital switchgear room will be modeling performed in support seals where the bus bars provided with penetration seals. of the fire PRA by preventing penetrate the floor of the impacts to equipment in the non-vital switchgear room adjacent room.

beneath the 1A-01, 1A-02, 2A-01, and 2A-02 CO!IJ[lensato!:)L measures for Buses. NFPA 805:

Appropriate compensatory measures are established as required until the modification is implemented.

Revision 0 Page S-17

ATTACHMENT D LAR UPDATES FOR PBNP RAt FPE 03

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.6 Standpipe and Standpipe and Hose N/A N/A- General statement; No N/A Hose Stations Stations. technical requirements use of EEE's 3.6.1 For all power block buildings, Complies with Clarification Standpipe and hose systems Fire Protection Technical Class Ill standpipe and hose were reviewed against the Evaluation (FPTE) 016, "Resolution systems shall be installed in requirements of NFPA of Identified Deviations to National accordance with NFPA 14, 14-1963 Edition in V878 Fire Protection Code "Standard for the Installation TD-05. Justification of Requirements," Rev. 6/ Sections of Standpipe, Private deviations from this code is 14-DEV-1 through 14-DEV-9, et al.

Hydrant, and Hose Systems." provided in FPTE 016, Sections 14-DEV-1 through NFPA 14, "Standard for the 14-DEV-9, et al. Installation of Standpipe and Hose Systems," 1963 Edition I All Technical Document V878 TD-05, "NFPA Code Compliance Assessment," Rev. 0 Revision 0 PageA-84

NextEra PBNP Attachment A- NEI 04-02 Table 8-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements!Guidance Compliance Statement Compliance Basis Reference Document use of EEE's 3.9.1 (2) NFPA 15, "Standard for Complies with Glarifioation Water spray systems were Fire Protection Technical Water Spray Fixed Systems reviewed against the Evaluation (FPTE) 016, "Resolution for Fire Protection" requirements of NFPA of Identified Deviations to National 15-1962 Edition in V878 Fire Protection Code TD~05. Justification of Requirements," Rev. 6/ Sections deviations from this code is 15DEV-1 through 15-DEV-9 provided in FPTE 016, Sections 15-DEV-1 through NFPA 15, "Water Spray Systems 15-DEV-9. for Fire Protection," 1962 Edition I All Technical Document V878 TD-05, "NFPA Code Compliance Assessment," Rev. 0 3.9.1 (3) NFPA 750, "Standard on N/A Water mist systems are not N/A Water Mist Fire Protection installed at PBNP.

Systems" 3.9.1 (4) NFPA 16, "Standard for the N/A Foam-water systems are not N/A Installation of Foam~Water installed at PBNP.

Sprinkler ~nd Foam-Water Spray Systems" Revision 0 Page A-98

NextEra PBNP Attachment A- NEI 04~02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.10.1(2) NFPA 12A, "Standard on Complies '*'*ith Clarification Halon systems were reviewed NFPA 12A, "Standard on Halon Halon 1301 Fire Complies with use of against the requirements of 1301 Fire Extinguishing Systems,"

Extinguishing Systems" NFPA 12A-1987 Edition in 1987 Edition I All EEE's V878-04-TD-05. Justification of deviations from this code is Technical Document V878 provided in FPTE 016, TD-05, "NFPA Code Compliance Sections 12A-DEV-1 and 12A- Assessment," Rev. 0, dated DEV-2. 3/24/99 Fire Protection Technical Evaluation (FPTE) 016, "Resolution of Identified Deviations to National Fire Protection Code Requirements," Rev. 6/ Sections 12A-DEV-1 and 12A-DEV-2 Complies with Use of FPEE 1999-015 justifies the Fire Protection Engineering EEEEs acceptability of Evaluation (FPEE) 1999-015, overpressurized halon "NFPA 12A Code Deviation-suppression system agent Halon Agent Storage Container containers. Pressures," Rev. 0 I All 3.10.1(3) NFPA 2001, "Standard on N/A There are no clean agent N/A Clean Agent Fire extinguishing systems at Extinguishing Systems" PBNP.

3.10.2 Operation of gaseous fire Complies No Additional Clarification Drawing 35105 E-44, "C900 suppression systems shall Annunciator Panel Terminations,"

annunciate and alarm in the Rev.8/AII control room or other constantly attended location identified.

Revision 0 Page A*104

ATTACHMENT E LAR UPDATES FOR PBNP RAI FPE 04

NextEra PBNP Attachment.A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.11.4(b) Conduits shall be provided Complies by previous Per Section 5.1.3.4 of the Fire Protection Evaluation with an internal fire seal that NRC approvaL FPER, *conduits that Report has an equivalent fire- penetrate fire barriers have (FPER), Rev. 13/ Section resistive rating to that of the been conservatively provided 5.1.3.4 fire barrier through opening with internal seals, either at fire stop and shall be the wall or at the first Conduit Fire Protection permitted to be installed on available access point of the Research Program Final either side of the barrier in a conduit. Small eonduits that Report, June 1, 1987 location that is as close to the are provided for items such barrier as possible. as lighting circuits and are Fire Protection Technical embedded in the concrete Evaluation (FPTE) 001, Exception: Openings inside construction are not "Technical Evaluation of Fire conduit 4 in. (1 0.2 em) or less considered to be paths for Barrier Penetration Seals, Fire in diameter shall be the spread of fire and have Rated Wrapping and Cable sealed at the fire barrier with a not been sealed. Tray Fire Stops at Point Beach fire-rated internal seal unless Nuclear Plant". Rev. 1 I the conduit extends greater Conduits that penetrate fire Section 4.3 than 5 ft (1.5 m) on each side barriers have been of the fire barrier. conservatively provided with Letter VPNPD-97-263 from Fay In this case the conduit internal seals, either at the (WE) to Craig (NRC) dated opening shall be provided with wall or at the first available June 19, 1987/ Submit Conduit noncombustible material to access point of the conduit. Fire Protection Research prevent the passage of smoke Small conduits that are Program Final Report for and hot gases. The fill depth provided for items such as Review of the material packed to a lighting circuns and are depth of 2 in. Letter NPC-30326 from (5.1 em) shall constitute an embedded in the concrete construction are not McCracken (NRC) to Fay (WE) acceptable smoke and hot gas dated May 18, 1988 I Request seal in this application. considered to be paths for the spread of fire and have not for Additional Information been sealed.

The general installation Letter VPNPD-88-325 from Fay criteria for installing internal (WE) to McCracken (NRC) conduit seals at PBNP dated June 15, 1988/

includes: Response to RAis Letter CLB-R0234 from

1. Conduits of two (2) inch McCracken (NRC) to Fay (WE) diameter or larger that dated October 23, 1989 /Issue terminate within an area Safety Evaluation Report shall be sealed at the first available fitting after penetrating the fire barrier. The seal shall consist of a noncombustible material such as "Kaowool" ceramic fiber, hand packed into the conduit or fittinQ to close any Page A-134.1

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document void space between the cable and the inner wall of the conduit. A conduit thus sealed shall then have the fitting cover put in place and painted, receive a penetration number and be entered into the penetration seal list and drawings as is currently done with silicone seals. Conduits less than two (2) inches in diameter, which terminate less than one (1 )footfrom the fire barrier shall be sealed in the same manner.

2. Conduits less than two (2)inches in diameter, which terminate one (1) foot or more from the fire barrier need not be sealed.

3. Conduits that run through a fire area but do not terminate in that area need not be sealed in that area.

In 1986 a group of 23 nuclear utilities sponsored by Wisconsin Electric (WE) embarked on a research project to evaluate internal conduit sealing requirements and develop design guidelines based on fire test data. This effort was initiated due to variations in conduit fire seal criteria guidance in Branch Technical Position (BTP) 9.5-1 Appendix A, NUREG 0800 and 10 CFR 50.48 A_l)Q_endix R and in Page A-134.2

ATTACHMENT F LAR UPDATES FOR PBNP RAI FPE 06

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document the general interest in providing cost effective fire protection based on sound technical information.

Results of this project were documented in a proprietary report entitled "Conduit Fire Protection Research Program." This report, prepared by "Professional Loss, Inc." was submitted to the NRC on June 19, 1987 for review. NRC issued the results of their review of Conduit Fire Protection Research Program in a Technical Evaluation Report (TER) and accompanying Safety Evaluation. The TER was prepared by technical assistance contractor Science Applications International Corporation (SAIC). Based on the staffs review, it was concluded that the subject research program justifies the implementation of revised conduit fire seal criteria.

Page A-134.3

NextEra PBNP Attachment A- NEI 04*02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document 3.3.5.3 Electric cable construction Complies by Previous Per Section A.4 of the Letter NPC-27492 from Burstein shall comply with a flame NRC Approval attachment to NPC-27492, (WE) to Lear (NRC) dated June 20, propagation test as "The 5 and 15kV cables were 1977 I Attachment, Section A.4 acceptable to the AHJ. required to pass the vertical flame-resistance test in Letter NPC-29597 from Schwencer accordance with IPCEA (NRC) to Burstein (NRC) dated S-19-81, Section 6.19.6. The August 2, 19791 Enclosure 3, safety related 600V power Safety Evaluation Report and control cables were (NPC-35989), Section 4.8 required to pass the vertical flame-resistance test in Fire Protection Technical accordance with IPCEA Evaluation (FPTE) 011, "Technical S-19-81, Section 6.19-6 and Evaluation of Acceptable PBNP the Bonfire Test..." Cable Fire Retardancy Standards Power, Control and Specialty Per Section 4.8 of Cable," Rev. 0 I All NPC-35989, "Safety-related electrical cables used at Point Beach were required to pass the vertical flame resistance tests in accordance with Insulated Power Cable Engineer's Association (IPCEA) test S-19-81, Section 6.19.6. Also, a test was conducted in which a bundle of jacketed cables was suspended with both ends approximately five inches above an open container of oil. The span was approximately five feet. The burning oil engulfed the cables for five minutes.

Cables are considered to pass the above flame tests if the material fails to propagate five minutes after the flame source is removed. The safety-related and nonsafety-related 600-Volt instrumentation cables w~re required to pass the IPCEA and oil fire tests. The safety-related 600-Volt power Revision 0

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document and control cables were-required to pass the IPCEA tests. The Institute of Electrical and Electronics Engineers (IEEE) test requirements were not in effect at that time. We find that retest to the IEEE 383 procedures and criteria would not provide information that would alter our recommendations or conclusions regarding the acceptability of the jacketing and insulation. Wherever there is concern that rapid flame propagation from an electrical fire vvill compromise redundant safety-related divisions, a qualified flame retardant coating or material will be used to minimize the likelihood. On this basis, we find the electrical cables used at Point Beach acceptable."

The electrical cables, as approved by the referenced SER, are still installed at F'BNP. There have been no plant modifications or other qhanges that would invalidate t!1e basis for approval.

Revision 0 Page A-21

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document Complies, with Required See implementation item Procedure DG-E13, "Insulated Action identified below. Electrical Cable Insulation Installation," Rev. 11 All NAMS Action Request 1855682, "Rev to Design Guideline DG-E 13 for Fire Retardant Coatings," dated March 12, 2013 IMPLEMENTATION ITEMS:

IMP-136: Design Guideline DG-E13, "Insulated Electrical Cable Insulation Installation," is being revised to require that, wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions, a qualified flame retardant coating or material shall be used to minimize the likelihood. This is being tracked by NAMS Action Request 1855682 make reference to FPTE 011, "Technical Evaluation of Acceptable PBNP Cable Fire Retardancy Standards Power, Control and Specialty Cable," which provides the acceptable cable qualification standards to meet the August 2, 1979 NRC commitment to use a qualified flame retardant coating or material wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions.

3.3.6 Roofs Metal roof deck construction Complies No Additional Clarification Fire Protection Evaluation Report shall be designed and (FPER), Rev. 131 Section 5.1.7 installed so the roofing system will not sustain a self- NFPA 256, "Standard Methods of propagating fire on the Fire Tests of Roof Coverings; underside of the deck when 1998 Edition I Section 1-1.2.1 the deck is heated by a fire inside the building. Roof Specification 6118-C-44, coverings shall be Class A "Specification for Furnishing and as determined by tests Installation of Built-up Roofing and described in NFPA 256, Insulation, Waterproofing and "Standard Methods of Fire Dampproofing," Rev. 1/ Section Tests of Roof Coverings." 10.2.1 Specification 6118-C-45, "Precast Concrete Roof Decking," dated May 1, 1968 I Section 6.0 Revision 0 Page A-22

Security-Related Information- Withhold from Public Disclosure Under 10 CFR 2.390 NextEra PBNP Attachment S - Modifications and Implementation Items Table S-3 Implementation Items Item Unit Description LAR Section I Source IMP-130 1,2 Pre-fire plans and fire brigade training materials will be revised to address radioactive 4.1.2 and Attachment A release requirements of NFPA 805. This is being tracked by NAMS Action Request 4.4 and Attachment E 1845926.

IMP-131 1,2 NFPA 805 radioactive release review will be incorporated into PBNP's configuration 4.4 and Attachment E control program. This is being tracked by NAMS Action Request 1845932.

IMP-132 1,2 Performance-based surveillance frequencies will be established for the turbine bearing 4.1.2 and Attachment A dry chemical suppression systems as described in Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide." Establishment of performance-based frequencies is being tracked by NAMS Action Request 1789890.

IMP-133 1,2 Performance-based surveillance frequencies will be established for fire dampers that do 4.1.2 and Attachment A not meet deterministic testing frequencies per NFPA 90A as described in Electric Power Research Institute (EPRI) Technical Report TR-1 006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide." Establishment of performance-based frequencies is being tracked by NAMS Action Request 1789890.

IMP-134 1,2 Procedure Ml 36.3, "Paint and Coatings," is being revised to require that wall and ceiling 4.1.2 and Attachment A finishes installed in the plant shall meet class A requirements of flame spread 0-25 and smoke development 0-450; and floor finishes shall meet class I requirements of critical radiant flux of not less than 0.45 W/cm2

• This is being tracked by NAMS Action Request 1855689. I IMP-135 1,2 Fire protection program documents will be updated and training will be provided as 4.7 necessary. This includes fire protection design basis document, system-level design basis documents and procedures, the Fire Protection Evaluation Report (FPER), the Fire Hazards Analysis Report (FHAR), the Safe Shutdown Analysis Report (SSAR), post-transition change process (including Fire PRA updates), and qualification training. This is being tracked by NAMS Action Request 1882226.

IMP-136 1,2 Design Guideline DG-E13, "Insulated Electrical Cable Insulation Installation," is being 4.1.2 and Attachment A revised to require that, wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions, a qualified flame retardant coating or material shall be used to minimize the likelihood. This is being tracked by NAMS Action Request 1855682 make reference to FPTE 011, "Technical Evaluation of Acceptable PBNP Cable Fire Retardancy Standards Power, Control and Specialty Cable," which provides the acceptable cable qualification standards to meet the August 2, 1979 NRC commitment to use a qualified flame retardant coating or material wherever there is concern that rapid flame propagation from an electrical fire will compromise redundant safety-related divisions.

Revision 0 Page S-27

ATTACHMENT G LAR UPDATES AND OTHER DOCUMENTS FOR PBNP RAI FPE 07

NextEra PBNP Attachment L- NFPA 805 Chapter 3 Requirements for Approval An unprotected rolling steel door, an unprotected single leaf swinging door, and a louver are present in the Turbine Building east wall opposite the hydrogen system at grade. Several louvers are present just below the roof line at an elevation of approximately 100 feet. Roll-up doors are present in the Turbine Building east wall above the hydrogen storage location.

A liquid nitrogen storage tank is located approximately 14.5 feet from the hydrogen system and is near the Turbine Building rolling steel door. The tank is surrounded on three (3) sides by guardrails standing approximately one (1) foot off the ground.

• Code Requirements Summary:

NFPA 567, Table 2 specifies the minimum separation distances of the hydrogen system from an outdoor exposure. The separation distances are based primarily on three criteria: size/volume of the hydrogen system, type of outdoor exposure, and the construction type of the outdoor exposure. For PBNP, the total hydrogen system volume is in excess of 15,000 scf; therefore the criteria for hydrogen quantity "in excess of 15,000 cubic feet" contained in Table 2 apply to this plant configuration.

The code requirements of Table 2 are summarized as follows:

o The Turbine Building is constructed of non-combustible concrete and sheet metal. Although the lower portion of the concrete wall is horizontally adjacent to the hydrogen system, the more restrictive separation distance criteria for "non-combustible construction" in Table 2 is used, as the higher portions of the wall is constructed of sheet steel. Therefore, the required separation distance between the Turbine Building and the hydrogen system is 25 feet.

o Table 2 also specifies that the Turbine Building wall openings, such as the doors and louver located~aLgr2de, be separated from the hydrogen system by a minimum 10 feet when the openings are not above any part of the hydrogen system. The Turbine Building east wall a-AS openings located at grade are approximately 35 feet away from the hydrogen system.

Basis for Request:

The basis for the approval request of this code deviation is as follows:

• The Turbine Building east wall, excluding openings, does not meet the required minimum separation requirements. However, based on analysis, the existing separation distance between the Turbine Building and hydrogen system is acceptable per field survey and review of the PBNP Fire Hazards Analysis Report (FHAR), plant drawings, and combustible loading analysis; which summarizes the following applicable hazards inside the east wall of the Turbine Building, identified as Fire Zones 547, 583, and 588, and the existing fire protection features within these zones:

o All fire zones along the east wall of the Turbine Building in the vicinity of the hydrogen storage area have low combustible loading; o Fire Zone 547 is equipped with portable fire extinguishers and hose reels; Revision 0 Page L-19

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