ML14079A159

Browns Ferry Units 1, 2, and 3, Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants

ML14079A159
Person / Time
Site:	Browns Ferry
Issue date:	03/14/2014
From:	Shea J W Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-14-025, L44 140314 001, TAC MF1185, TAC MF1186, TAC MF1187
Download: ML14079A159 (61)

Text

Security-Related Information

-Withhold f r om Public Disclosure in accordance with 10 CFR 2.390. Attachment 2 of Enclosure 1 contains Security-Related Information.

Upon removal of Attachment 2 from Enclosure 1 , this letter is uncontrolled. L44 140314 001 1101 Market Street , Chattanooga, Tennessee 37402 CNL-14-025 March 14,2014 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2 , and 3 10 CFR 50.90 10 CFR 2.390 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260 , and 50-296

Subject:

Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187) -Set 5 References

1. Letter from TVA to NRC, "License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (Technical Change TS-480)," dated March 27, 2013 (ADAMS Accession No. ML 13092A393)

2. Letter from TVA to NRC , "Response to NRC Request to Supplement License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2 , and 3 (TAC Nos. MF1185, MF1186 , and MF1187)," dated May 16 , 2013 (ADAMS Accession No. ML 13141A291)

3. Letter from NRC to TVA , "Browns Ferry Nuclear Plant, Units 1 , 2 , and 3 -Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (TAC Nos. MF1185, MF1186 , and MF1187)," dated November 19,2013 (ADAMS Accession No. ML 13298A702)

U.S. Nuclear Regulatory Commission Page 2 March 14, 2014

4. Letter from TVA to NRC, "Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187) - Set 2," dated January 10, 2014 (ADAMS Accession No. ML1401A088)

5. Letter from TVA to NRC, "Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187) - Set 4," dated February 13, 2014 (ADAMS Accession No. ML14055A305)

By letter dated March 27, 2013 (Reference 1), Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, to transition to National Fire Protection Association Standard (NFPA) 805. In addition, by letter dated May 16, 2013 (Reference 2), TVA provided information to supplement the Reference 1 letter.

By letter dated November 19, 2013 (Reference 3), the Nuclear Regulatory Commission (NRC) requested additional information to support the review of the LAR. The required dates for responding to the requests for additional information (RAIs) varied from a nominal

60 days to 120 days.

provides the fifth set of TVA responses to some of the RAIs identified in the Reference 3 letter. This enclosure provides the remainder of the nominal 120 day responses that were not previously submitted. These nominal 120 day responses are due by March 15, 2014. In addition, Attachment 1 to Enclosure 1 includes markups of the LAR pages that are affected by the TVA response to FPE RAI-11. These LAR changes cannot be easily described in the response; therefore markup pages are provided. Furthermore, to Enclosure 1 contains security-related information and should be withheld from public disclosure under 10 CFR 2.390.

provides updated TVA responses to FM RAI 03, Part b.i, which was previously responded to in the Reference 4 letter, and PRA RAI 01.h, Part ii, which was previously responded to in the Reference 5 letter.

Enclosure 3 provides a listing of the RAIs listed in the Reference 3 letter and the actual date of the TVA response to each of the RAIs.

provides a discussion of the LAR Attachment S, Table S-2 committed modifications that TVA is either modifying or deleting, consistent with the discussion held on February 20, 2014 with the NRC BFN Project Manager.

U.S. Nuclear Regulatory Commission Page 3 March 14 , 2014 Consistent with the standards set forth in Title 10 of the Code of Federal regulations (10 CFR), Part 50.92(c), TVA has determined that the additional information , as provided in this letter , does not affect the no significant hazards consideration associated with the proposed application previously provided in Reference

1. There are no new regulatory commitments contained in this submittal.

Please address any questions regarding this submi t tal to Mr. Edward D. Schrull at (423) 751-3850. I declare under of perjury that the foregoing is true and correct. Executed on this 14th day of March 2014. President , Nuclear Licens i ng Enclosures

1. TVA Responses to NRC Request for Additional Information

Set 5 (nominal 120-day) 2. Updated TVA Response to NRC Request for Additional Information FM 03 , Part b.i and PRA 01.h, Part ii 3. Summary of BFN NFPA 805 RAI Response Dates 4. Revisions to Proposed Modifications cc (Enclosures)

NRC Regional Administrator

-Region" NRC Senior Resident Inspector

-Browns Ferry Nuclear Plan t NRC Project Manager -Browns Ferry Nuclear Plant State Health Officer, Alabama Sta t e Department of Health E1-1 ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 TVA Responses to NRC Request for Additional Information: Set 5 (nominal 120-day)

FPE RAI 10 LAR Attachment S, Table S-2 identifies proposed installation of incipient detection system modifications. For the proposed installation of very early warning fire detection systems, provide more details regarding NFPA code(s) of record (including year), proposed installation configuration (common piping or individual cabinet or area wide), acceptance testing, sensitivity and setpoint control(s), alarm response procedures and training, and routine inspection, testing, and maintenance that will be implemented. If the system(s) have not yet been designed or installed, provide the specified design features for the proposed system. The specified design features should include the installation testing criteria to be met prior to operation. Because

Frequently Asked Question (FAQ) 08-0046 (ADA MS Accession No. ML093220426) is identified in the LAR Attachment H, Table H-1, describe whether this design and installation will be in compliance with each of the elements, limitations and criteria of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," Chapter 13, and FAQ 08-0046 including the closeout memo. Provide justification for any deviations.

RESPONSE:

LAR Attachment S, Table S-2 identifies three modifications associated with incipient detection systems (i.e., very early warning fire detection systems). These three modifications are Items 3, 77, and 78.

LAR Attachment S, Table S-2, Item 77 proposes modifications to install area wide incipient detection (i.e., a very early warning fire detection system) in the auxiliary instrument rooms (Fire Compartments 16-K, 16-M, and 16-O). TVA has decided not to install the area wide very early warning fire detection system in the auxiliary instrument rooms. This decision was made because the risk increase of removing the credit for prompt detection is less than 1% total plant Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Therefore, the effect on the Fire PRA is negligible. LAR Table S-2 is revised to delete Item 77. In addition, LAR Section V.2.4 is revised to remove discussions of credit for area wide very early warning fire detection in the auxiliary instrument room (see Attachment 1 of this enclosure). The TVA response to FPE RAI 12, Part c provides further information on the effect of removing the area wide very early warning fire detection system from the Fire Probabilistic Risk Assessment (PRA). In-cabinet protection will still be provided.

The detection systems identified in LAR Attachment S, Table S-2, Item 78 will provide area wide coverage for the Units 1, 2, and 3 Cable Sp reading Rooms (CSRs).

These systems have not been designed and are addressed in the TVA response to FPE RAI 12, Part d (in this enclosure). A discussion of the current conceptual design approach is also provided in the TVA response to FPE RAI 12, Part d.

LAR Attachment S, Table S-2, Item 3 addresses incipient detection to be installed in the electrical panels in the Units 1, 2, and 3 Auxiliary Instrument Rooms. These detection systems have been designed and the modifications are being implemented. These detection systems are addressed in the remainder of this RAI response.

E1-2 NFPA 72-2010, "National Fire Alarm and Signaling Code," applies to these detection systems along with meeting the requirements of NFPA 76-2012, "Standard for the Fire Protection of Telecommunications Facilities," for response transport times and sensitivity settings. Code compliance reviews will be performed as part of implementation.

The design of the detection systems will include one detector installed in each Auxiliary Instrument Room, with each detector monitoring four zones. Each of the four zones will monitor the electrical panels in one of the four rows of panels. The detection system will send an alarm to the main control room (MCR) fire alarm annunciator and Fire Operation's annunciator. This would alert personnel to respond to a potential fire so it can be extinguished manually during the incipient stage. The system will indicate which of the four zones are in alarm to permit personnel to investigate the row of panels from which the alarm is originating.

Testing and commissioning of each incipient detector will be completed in accordance with the vendor's acceptance test and associated sensitivity testing. The vendor commissioning of the detector demonstrates compliance with criteria established by applicable standards, which includes testing the sensitivity and transport time. In accordance with NFPA 76, this type of system is required to have a transport time of no greater than 60 seconds from any one sampling point. The sensitivity and setpoints will be controlled by surveillance procedure(s). Routine inspection, testing and maintenance will be conducted in accordance with vendor recommendations, including sensitivity and transport time tests.

Procedures and training will be developed as part of NFPA 805 implementation covering responses to an alarm. Personnel will respond to alarm conditions to locate the source and extinguish any fire that may occur.

Design and installation will be in compliance with each of the elements, limitations and criteria of

NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," Chapter 13, and FAQ 08-0046 including the closeout memo (ADAMS Accession No. ML093220426) with no deviations.

E1-3 FPE RAI 11

Incipient detection is described in the LAR in a variety of configurations, credits, and locations.

a) During the audit, the NRC staff observed the installation of incipient detection to monitor electrical panels in the auxiliary instrument rooms for risk reduction. LAR Attachment S, Table S-2, Item 77 proposes to install area wide incipient detection in the same rooms for transient fires. Provide a more detailed description of how these systems will be designed, installed, operated, and maintained.

b) LAR Attachment S, Table S-2, Items 78 and 79 propose to install incipient detection that will actuate a total flooding clean agent suppression system for transient fires. Provide a more detailed description of these systems, and how are they credited in the FPRA. If area wide incipient detection is intended to be credited in the FPRA, provide justification for this type of application within the context of the current NUREG/CR-6850 and FAQ 08-0046 as referenced in LAR, Attachment H, Table H-1.

c) Facts and Observations (F&O) 2-57 in LAR Attachment V, Table V-7 indicates that "incipient fire detectors" are installed to monitor electrical cabinets in physical analysis units (PAUs) (e.g., fire compartment): 16-M, 16-O, and 16-K. Provide a description of the incipient detection system(s) being relied upon including whether each system is in-cabinet or area wide.

RESPONSE:

Part a The area wide incipient detection systems (i.e., area wide very early warning detection systems) proposed for Fire Compartments 16-K, 16-M, and 16 O listed in LAR Attachment S, Table S-2, Item 77 will not be installed in the Unit Auxiliary Instrument Rooms. Therefore, these detection systems will not be addressed in this RAI response. See the TVA response to FPE RAI 10 (in this enclosure) for additional information.

Part b LAR Attachment S, Table S-2, Modifications 78 and 79 propose modifications to install an area wide very early warning fire detection system and a new automatic gaseous fire suppression system in the CSRs. Although the designs for these modifications are still conceptual, the design details for the proposed area wide very early warning fire detection system and the total flooding clean agent system in the CSR portion of Fire Compartment 16-A, are provided in the TVA responses to FPE RAI 10 (Modification 78 only) and FPE RAI 12, Part d (Modifications 78 and 79).

As discussed in LAR Attachment V,Section V.2.4, the Fire PRA applies a prompt detection credit and an automatic suppression credit for transient fires, cable fires caused by welding and cutting, transient fires caused by welding and cutting, and self-ignited cable fires.

The Fire PRA credit for the area wide detection system is discussed in detail in the TVA response to PRA RAI 10, Part a. With the exception of the sampling point locations (i.e., area wide versus in-cabinet), the Fire PRA credit was applied consistent with the guidance from NUREG/CR-6850 and NUREG/CR-6850, Supplement 1 (i.e., FAQ 08-0046) and is justified as follows:

E1-4 Both area wide and in-cabinet very early warning fire detection systems are required to be designed and installed to the same codes and standards (i.e., NFPA 72 and NFPA 76) and are required to be listed by a nationally recognized testing laboratory (i.e., Underwriter's Laboratory (UL)). The appropriate manual fire suppression failure probabilities in NUREG/CR-6850, Supplement 1 (i.e., FAQ 08-0050) are selected based on the specific initiator. The unavailability and unreliability of the area wide very early warning fire detection system is considered equivalent to an in-cabinet system, as further discussed in the TVA response to FPE RAI 12, Part a. The detection system unavailability and unreliability value of 1E-02 from FAQ 08-0046 is appropriately included. Prompt detection has been credited for the subject CSR fire scenarios in accordance with the guidance from NUREG/CR-6850, A ppendix P for a high-sensitivity smoke detection system. Even when the detection system and manual s uppression, via the fire watch for welding and cutting scenarios, are successful, the analysis still assumes damage to one cable tray. The tray damaged by the fire is conservatively assumed to be one of the top 25 risk contributing cable trays in Fire Compartment 16-A for each unit. Fire scenario end states for unsuccessful detection or suppression conservatively assume all targets in the respective unit's CSR are damaged (i.e., whole room damage).

Part c LAR Attachment S, Table S-2, Modification 3 proposes modifications to install in-cabinet incipient detection to monitor the electrical panels in each unit's auxiliary instrument room.

Specific design information for the incipient detection systems being installed to monitor electrical cabinets in physical analysis units (PAUs) (i.e., fire compartment) 16-M, 16-O, and 16-K is provided in the TVA response to FPE RAI 10. The incipient fire detection systems installed in Fire Compartments 16-M, 16-O and 16-K are in-cabinet only.

E1-5 FPE RAI 12

LAR Attachment V,Section V.2.4 applies an area-wide "prompt" detection credit and an automatic suppression credit for transient fires; cable and transient fires caused by welding and cutting; and self-ignited cable fires in the Cable Spreading Rooms (CSRs). Justification for such credit is listed. Provide the following additional information:

a) The LAR indicates the use of FAQ 08-0046 availability and reliability values for the area-wide detection credit. However, this FAQ only addresses in-cabinet detection. Provide justification for this implied equivalency of area wide and in-cabinet detection.

b) Provide a more specific description of the incipient detection system in "Fire Compartment 16A." Include the type of system(s) and location of detectors (area wide vs. in-cabinet), especially with regard to the "top 25 risk contributing cable trays."

c) The FPRA also applies a prompt detection credit for two transient fire scenarios in the

Unit 1 auxiliary instrument room (Fire Area 16/Fire Compartment 16-K). Describe how prompt detection credit is determined.

d) Describe how the incipient detection system will be used to actuate the automatic suppression system. Describe whether a pre-action alarm and time delay prior to actuation will be incorporated in the design, and if so, describe how this would be factored into the FPRA.

RESPONSE:

LAR Attachment S, Table S-2, Modifications 78 and 79 propose modifications to install, in the CSRs, an area wide incipient (i.e., very early warning) fire detection system, that uses very early warning fire detectors (VEWFDs), and a new automatic gaseous fire suppression system. As discussed in the TVA response to FPE RAI 10, TVA has determined that the proposed modification described in LAR Attachment S, Table S-2, Modification 77 to install area wide very early warning fire detection systems in the auxiliary instrument rooms (i.e., Fire Compartments 16-K, 16-M, and 16-O), and discussed in LAR Attachment V,Section V.2.4, will not be installed.

Part a The CSR area wide very early warning fire detection system will actuate a total flooding gaseous suppression system. Additional design information for these systems and further details regarding how the detection system will be used to actuate the automatic suppression system are provided in the TVA response to FPE RAI 12, Part d.

The availability and reliability of the CSR area wide very early warning fire detection system is equivalent to in-cabinet very early warning fire detection systems based on the following:

Both area wide and in-cabinet very early warning fire detection systems are required to be designed and installed to the same codes and standards (i.e., NFPA 72 and NFPA 76) and are required to be listed by a nationally recognized testing laboratory (i.e., UL). The obscuration criteria and location of detectors (i.e., pipe openings) for area wide detection systems will follow NFPA 72 standards, NUREG/CR 6850, Supplement 1, Section 13.2, and manufacturer listings. NFPA 76, as referenced in FAQ 08-0046, provides guidance on area wide application of the detection systems.

E1-6 Area wide detection systems will be operated in accordance with referenced standards such as NFPA 76 and NFPA 72, and manufacturer requirements, that are the equivalent to the standards regulating the in-cabinet detection systems. The air sampling detectors will follow routine maintenance and inspection procedures to ensure optimum reliability and availability of the detector. Availability of the area wide systems is similar to in-cabinet protection methods as the same protocols will apply in terms of maintenance and monitoring. The detection system will be supervised for multiple troubles; failures will be indicated by a trouble signal on the fire alarm system that reports to the MCR fire alarm annunciator and Fire Operation's annunciator. If area wide fire detection is unavailable, then compensatory measures will be provided, consistent with the methodology for in-cabinet detection systems.

Part b Although the designs for the modifications installing the area wide very early warning fire detection system and automatic gaseous fire suppression system in the CSR portion of Fire Compartment 16-A are still conceptual, the following discussion provides the current modification approach.

The new detection system will also be used to activate the clean agent gaseous suppression systems. The gaseous suppression systems w ill provide area wide coverage for the entire CSR, regardless of the location of the top 25 risk contributing cable trays.

Specific design and plant operation information is provided in the TVA response to Part d of this RAI, for the detection and gaseous suppression systems.

The TVA response to Part d of this RAI also provides additional design details for the detection system, including the type of system, location of sampling points, and how the detection system will actuate the automatic suppression system.

Part c The prompt detection credit was taken for two transient fire scenarios in the Unit 1 auxiliary instrument room (i.e., Fire Compartment 16-K) based on the proposed plant modification for the installation of an area wide very early warning fire detection system in the fire compartment (i.e., LAR Attachment S, Table S-2, Modification 77). This modification will not be installed, as discussed in the initial paragraph of this response. Therefore, prompt detection in Fire Compartment 16-K will no longer be credited for transient fire scenarios.

A sensitivity analysis has been performed to analyze the effect of removing the prompt detection credit (i.e., by removing suppression from the transient fire scenarios). The risk increase results from the removal of the prompt detection credit for two transient fire scenarios in Fire Compartment 16-K, are provided in Table FPE RAI 12.c, in Attachment 2 to this enclosure.

The risk increase of removing the credit for prompt detection is less than 1% total plant Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Therefore, the effect on the Fire PRA is negligible. The Fire PRA will be updated to remove the credit for the area wide fire detection system in Fire Compartment 16-K. LAR Attachment W will be revised to reflect the results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to the remaining NRC RAIs.

E1-7 Part d Although the proposed designs for the modifications installing the area wide detection system and automatic gaseous fire suppression system in the CSR portion of Fire Compartment 16-A are still conceptual, the following discussion provides the current modification approach.

The detection system will be designed to locate sampling points for area-wide detection to provide optimum coverage for the detection scheme. The detection system will be capable of sending a signal to actuate a total flooding clean agent suppression system. The detection system will be designed and installed in accordance with NFPA 72-2010 and will meet the requirements of NFPA 76-2012 for response transport times and sensitivity settings. The clean agent suppression system will be designed in accordance NFPA 2001-2012.

The detection system will have alarm thresholds, where the lower level alarm thresholds will initiate a signal to the plant fire alarm system. These lower level alarm thresholds will not initiate activation of the clean agent suppression system. A higher level alarm threshold will actuate the clean agent suppression system for the room being monitored. The design of the suppression system will incorporate adequate gaseous concentrations to perform the required suppression function. Nozzles will be located to be capable of properly extinguishing a fire in the CSRs. There will be no time delay factored into the activation of the clean agent suppression system once the highest alarm threshold is achieved.

The TVA response to PRA RAI 10 discusses how the designs are treated in the Fire PRA.

E1-8 FPE RAI 13

In LAR Attachment I, Table I-1 lists power block structures and identifies corresponding plant fire areas, but does not correspond with other LAR tables. The following need additional clarification:

a) The following three fire areas: off gas building (OG), standby gas treatment building (SGTS), cooling towers (CT) are included in LAR Attachment W, Tables W-8, W-9, and W-10, but listed as part of the YARD in Table I-1. LAR Attachment C, Table C-2 doesn't list these fire areas either. Provide clarification for these apparent discrepancies.

b) Describe how each structure in the YARD fire area are differentiated (e.g., core damage

frequency (CDF) and Delta () CDF by Unit) between Tables W-8, W-9, and W-10.

c) In LAR Attachment W, Tables W-8, W-9, and W-10, the following three fire areas: off gas, standby gas treatment building, cooling towers are identified with Note 1: "Listed for PRA boundary." Define "Listed for PRA boundary" and indicate whether it was considered within the FPRA. Describe why these areas are marked "Not Applicable" for FREs. RESPONSE:

Part a The Off Gas Building (OG), Standby Gas Treatment Building (SGTS), and Cooling Towers (CT) are all Physical Analysis Units (PAU) within Fire Area "YARD." These structures are correctly identified as compartments within Fire Area "YARD" in LAR Attachment I, Table I-1. LAR Attachment W, Tables W-8, W-9, and W-10 have been clarified, as discussed in the TVA responses to Parts b and c below, to include OG, SGTS, and CT as being part of Fire Area "YARD."

Part b The OG, SGTS, and CT are power block structures contained within the YARD fire area and were listed in LAR Attachment W, Tables W-8, W-9, and W-10 as separate entries. Note 1 to the three tables was meant to explain that they were part of the Fire Area Yard, but were being listed separately in the three tables. However, for clarity and consistency with LAR Attachment I, Table I-1, these compartments and their respective risk, will be included in the YARD fire area (i.e., combined into one entry). Because there is zero delta risk for these areas, the delta risk for the YARD will not change.

The separate entries for OG, CT, and SGTS along with the note stating "Note 1: Listed for PRA boundary" is removed from LAR Attachment W, Tables W-8, W-9, and W-10, and the YARD entry is updated by summing the risk of the YARD, OG, SGTS, and CT. The specific values of the changes for the three units, i.e., Tables W-8, W-9, and W-10, are provided in FPE RAI 13 in to this enclosure.

Part c Note 1 to the three tables was meant to explain that the three compartments were part of the Fire Area Yard, but were being listed separately in the three tables. However, as discussed in the TVA response to Part b above, the OG, SGTS, and CT will be included in the YARD Fire Area and deleted as separate line items in LAR Attachment W, Tables W-8, W-9, and W-10. In E1-9 addition, the Fire Risk Evaluation column for these three compartments was marked as "NA" because there is zero delta risk for these three compartments (as explained in the TVA response to Part b above).

E1-10 FM RAI 01.i.iv

NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

The algebraic equations implemented in FDTs [Fire Dynamics Tools] and Fire Induced Vulnerability Evaluation, Revision 1 (FIVE) were used to characterize flame radiation (heat flux), flame height, plume temperature, ceiling jet temperature, and hot gas layer (HGL) temperature.

The Consolidated Model of Fire and Smoke Transport (CFAST) was used in the multi-compartment analysis (MCA), and for the temperature sensitive equipment hot gas layer study.

Fire Dynamics Simulator (FDS) was used to assess the MCR habitability, and in the plume/hot gas layer interaction and temperature sensitive equipment ZOI studies.

Section 4.5.1.2, "Fire PRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V, [Validation & Verification]" for a discussion of the acceptability of the fire

models that were used.

Regarding the acceptability of the PRA approach, methods, and data:

i. Specifically regarding the use of FDS in the MCR abandonment calculations:

iv. Due to computational constraints, a mesh size of 20 centimeter was used in the FDS MCR simulations. Consequently, for most scenarios the corresponding grid resolution (D

• /x) value was below the range of 4-16 suggested in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications". Provide technical justification for running FDS with D

• /x values outside the NUREG-1824 recommended range.

RESPONSE:

The 20 centimeter (cm) mesh size for the fire models using FDS in the MCR analysis was selected in order to maximize accuracy within a reasonable computation time. The majority of fire scenarios were not modeled with a large enough Heat Release Rate (HRR) to allow for the

D*/x to fall within the 4-16 range. By not being in range, the effects of the fire on targets in close proximity may not be as precise due to the larger mesh size. However, the main objective of the FDS models in the MCR analysis is to determine the time to operator abandonment by calculating the temperature, smoke concentration, radiant heat flux, and the hot gas layer throughout the entire compartment as opposed to damage of a single target. Although a smaller mesh size would allow for more precision, the selected mesh size was sufficient due to the large size of the control rooms.

A sensitivity analysis was performed to justify the 20 cm mesh size in the FDS models by changing the mesh size for multiple FDS models from 20 cm to 10 cm. For each FDS model analyzed as part of this sensitivity analysis, the updated D

• /x values were within the 4-16 range. This sensitivity analysis analyzed each type of ignition source postulated in the MCR (i.e., electrical cabinet and transient fires, inside and outside the horseshoe, in the Units 1 and 2 MCR and Unit 3 MCR). For each type of ignition source the fire scenarios with the largest E1-11 HRRs (i.e., Bin 15), were analyzed. In addition, smaller fires (i.e., Bin 5 and Bin 7) were also analyzed. The original abandonment times determined using the 20 cm mesh and the abandonment times determined using the 10 cm mesh are provided in the following table.

Abandonment Times for FDS model with 20 cm and 10 cm mesh sizes MCR Ignition Source Type and Location Bin Abandonment Time (minutes)

Abandonment Ratio (t 10cm/t 20cm) 20 cm mesh 10 cm mesh Units 1 and 2 Electrical Cabinet Fire Inside Horseshoe 15 8.5 8.2 0.96 Units 1 and 2 Electrical Cabinet Fire Outside Horseshoe 15 6.6 7.2 1.09 Units 1 and 2 Electrical Cabinet Fire Outside Horseshoe 7 9.5 10.5 1.11 Units 1 and 2 Transient Fire Inside

Horseshoe 15 7.7 7.3 0.95 Units 1 and 2 Transient Fire Inside

Horseshoe 5 14.8 14.1 0.95 Units 1 and 2 Transient Fire Outside

Horseshoe 15 7.3 6.2 0.85 Unit 3 Electrical Cabinet Fire Inside Horseshoe 15 6.2 6.2 1.00 Unit 3 Electrical Cabinet Fire Outside Horseshoe 15 6.3 6.5 1.03 Unit 3 Electrical Cabinet Fire Outside Horseshoe 7 8.5 8.6 1.01 Unit 3 Transient Fire Inside

Horseshoe 15 5.7 5.5 0.96 Unit 3 Transient Fire Inside

Horseshoe 5 9.0 9.0 1.00 Unit 3 Transient Fire Outside

Horseshoe 15 5.9 5.5 0.93

The abandonment ratio represents the amount of change expected for a specific type of ignition source. An abandonment ratio greater than 1 shows that the abandonment time calculated using the 10 cm mesh size was greater than that calculated using the 20 cm mesh size. Comparing the abandonment ratio for the large fires (i.e., Bin 15) to the smaller fires (i.e., Bin 5 or 7) shows a general consistency within each type of fire. As such, the abandonment ratio was determined to be a good representation of the change in abandonment time for each type of ignition source that was not reanalyzed during this sensitivity analysis. In addition, three of the four larger fire scenarios bounded the results of the smaller fire, so the larger fires were chosen to represent the bounding case. Therefore, the Bin 15 abandonment ratio was used to update the abandonment times in the sensitivity analysis for all bins for the applicable type of ignition source.

The abandonment times from the sensitivity analysis were used to recalculate the probability of non-suppression and the resulting probability of abandonment for each scenario. A comparison of the original abandonment probabilities and the abandonment probabilities determined as part of the sensitivity analysis are provided in the following table.

E1-12 Overall Abandonment Probabilities MCR Original Analysis Sensitivity Analysis Delta Units 1 and 2 2.07E-02 2.16E-02 8.64E-04 Unit 3 5.15E-02 4.75E-02 -4.11E-03 Total MCR Abandonment Probability (Units 1 and 2 MCR and Unit 3 MCR) 7.22E-02 6.90E-02 -3.18E-03

By comparing the original abandonment times to the results of the sensitivity analysis, the total probability of abandonment is shown to decrease if the mesh size for each scenario was reduced from 20 cm to 10 cm. Therefore, the abandonment ratio is accurately represented with the use of the 20 cm mesh size.

E1-13 FM RAI 02.a

American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/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 structures, systems, and

components. 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 flux for thermoset and thermoplastic cables as described in NUREG/CR-6850.

RESPONSE:

NUREG/CR-6850, Appendix H, provides damage and ignition criteria for both thermoset and thermoplastic cables that are typically found in nuclear power plants. Specifically, Table H-1 provides the damage criteria for both the radiant heat flux and temperature for both types of cables.

Raceways were conservatively analyzed as thermoplastic targets in all fire compartments at BFN unless the use of thermoset damage criteria is technically justified. In accordance with

NUREG/CR-6850, Appendix H, the following generic screening damage criteria were used for thermoplastic targets:

Critical Temperature: 205ºC (400ºF) Critical Heat Flux: 6 kW/m² (0.5 BTU/ft²s)

In accordance with NUREG/CR-6850, Appendix H, the following generic screening damage criteria were used for thermoset targets:

Critical Temperature: 330ºC (625ºF) Critical Heat Flux: 11 kW/m² (1.0 BTU/ft²s)

E1-14 FM RAI 02.b

American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/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 structures, systems, and

components. Appropriate temperature and critical heat flux criteria must be used in the analysis.

b. The technical documentation supporting the LAR that describes the fire modeling that was performed seems to imply that Institute of Electrical and Electronics Engineers (IEEE)-383, "IEEE Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations," qualified cables are assumed to be equivalent in terms of damage thresholds to "thermoset" cables as defined in Table 8-2 of NUREG/CR-6850. In addition, non-IEEE-383 q ualified cables are assumed to be equivalent to "thermoplastic" cables as defined in Table 8-2 of NUREG/CR 6850. These assumptions may or may not be correct. An IEEE-383 qualified cable may or may not meet the criteria for a "thermoset cable" as defined in NUREG/CR-6850. It is also possible that a non-IEEE-383 qualified cable actually meets the NUREG/CR-6850 criteria for a "thermoset" cable. Provide clarification on the assumptions that were made in terms of damage thresholds of cables.

RESPONSE:

All raceways, with the exception of cable trays in Fire Compartments (FCs) 05 and 09, were

conservatively analyzed as thermoplastic targets. In accordance with NUREG/CR-6850, Appendix H, the following thermal damage criteria have been used for thermoplastic targets:

Critical Temperature: 205ºC (400ºF) Critical Heat Flux: 6 kW/m² (0.5 BTU/ft²s)

The damage criteria for thermoset cables are:

Critical Temperature: 330ºC (625ºF) Critical Heat Flux: 11 kW/m² (1.0 BTU/ft²s)

Although IEEE-383 qualified cable may or may not meet the criteria for a "thermoset cable," modeling both IEEE-383 qualified and non-IEEE qualified cables based on thermoplastic damage criteria is conservative.

Within FC 05 and 09, cable trays were analyzed using the thermoset damage criteria. All other targets within FC 05 and FC 09 are assumed to be damaged with the thermoplastic damage thresholds.

A review of the cable insulation and cable jacket material was performed for the cables routed through cable trays in FC 05 and FC 09. The results of this review determined that all cables routed in cable trays within FC 05 have a thermoset insulation and jacket material. Therefore the use of thermoset damage criteria for cable trays in this fire compartment is valid.

The results of the cable jacket and insulation review in FC 09 determined that one cable tray contained cables with thermoplastic jacket material. Due to the conservative approach used for the fire modeling analysis for this fire compartment, each fire scenario that could damage the E1-15 cable tray has assumed damage to all cables in the cable tray at time equal 0. This bounds the reduced damage threshold for thermoplastic cables.

As discussed above, targets were analyzed using thermoplastic damage criteria in most fire compartments. Where thermoset damage criteria were used, the cable material type was confirmed as thermoset or, in the case of FC 09, the target damage set is bounding. Therefore, the current target damage thresholds analyzed are acceptable.

E1-16 PRA RAI 01.f

Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA is also referred to as PRA) approach, methods, and data shall be acceptable to the AHJ, which is the NRC. Regulatory Guide (RG) 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA 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. RG 1.200, "An Approach For Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated ASME/ANS 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 for evaluations that could influence the regulatory decision. The primary results of a peer review include the F&Os identified by the peer review and their subsequent resolution.

Clarify the following dispositions to fire F&Os and Supporting Requirement (SR) assessments identified in LAR Attachment V that have the potential to impact the FPRA results and do not

appear to be fully resolved:

f. F&O 2-54 against FSS-A6:

The disposition to this F&O states that the fire ignition frequency of individual main control board (MCB) scenarios is determined by dividing the total plant MCB fire ignition

frequency by the total number of unscreened MCB scenarios; however, use of the approach outlined in Appendix L of NUREG/CR-6850 requires the full Bin 4 frequency to be applied to each postulated MCB scenario. Provide the results of a sensitivity study (i.e., CDF, large early release frequency (LERF), CDF and LERF) that applies the guidance in Appendix L.

RESPONSE:

A sensitivity analysis that applied the guidance in NUREG/CR-6850, Appendix L was performed to analyze the risk impact that apportioning the full Bin 4 frequency (i.e., 8.24E-04) to each postulated Main Control Board (MCB) scenario would have on the total fire risk of the Post-Transition (PT) and Compliant (Comp) models for each unit. The location weighting factor was determined to be 1.5, following the guidance in NUREG/CR-6850, Table 6-2. The full Bin 4 frequency multiplied by the location weighting factor (i.e., 1.24E-03) was applied to each MCB fire scenario. All other values needed to calculate the CDF and LERF (e.g., Probability of Target Damage, Conditional Core Damage Probability (CCDP), Conditional Large Early Release Probability (CLERP)) for the MCB fire scenarios were not affected. The results of this sensitivity analysis are provided in Table PRA RAI 01.f in Attachment 2 to this enclosure.

The results of this sensitivity show a slight increase to the CDF and LERF for each Unit. Despite the increase, BFN meets the guidance for a Region II plant with total CDF and LERF below 1E-04/rx-yr and 1E-05/rx-yr, respectively, for ov erall plant risk. BFN also meets the change in CDF/LERF criteria for a Region II plant which allows a positive delta () CDF of 1E-05/rx-yr and LERF of 1E-06/rx-yr for acceptable risk increases.

E1-17 PRA RAI 01.o

Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA is also referred to as PRA) approach, methods, and data shall be acceptable to the AHJ, which is the NRC. Regulatory Guide (RG) 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA 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. RG 1.200, "An Approach For Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated ASME/ANS 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 for evaluations that could influence the regulatory decision. The primary results of a peer review include the F&Os identified by the peer review and their subsequent resolution.

Clarify the following dispositions to fire F&Os and Supporting Requirement (SR) assessments identified in LAR Attachment V of the license amendment request that have the potential to impact the FPRA results and do not appear to be fully resolved:

o. F&O 4-30 against FQ-D1:

The disposition to this F&O states that a dependency analysis was not performed between non-LERF and LERF actions. In addition, Section 5.9 of Post-Fire HRA report indicates that LERF actions are not included in the dependency analysis. Perform an HRA dependency analysis that considers all actions, including those for LERF, or alternatively provide a sensitivity analysis of the impact on risk estimates (i.e., CDF, LERF, CDF, and LERF) of including this dependency analysis. If actions are indeed independent, provide justification on a combination basis.

RESPONSE:

TVA will modify the Human Reliability Analysis (HRA) to consider LERF Human Failure Event (HFE) dependencies.

After this change in method is implemented, the affected sections in the LAR will be revised to reflect results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to the remaining NRC RAIs.

E1-18 PRA RAI 01.v

Section 2.4.3.3 of NFPA 805 states that the probabilistic safety assessment (PSA is also referred to as PRA) approach, methods, and data shall be acceptable to the AHJ, which is the NRC. Regulatory Guide (RG) 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA 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. RG 1.200, "An Approach For Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated ASME/ANS 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 for evaluations that could influence the regulatory decision. The primary results of a peer review include the F&Os identified by the peer review and their subsequent resolution.

Clarify the following dispositions to fire F&Os and Supporting Requirement (SR) assessments identified in LAR Attachment V that have the potential to impact the FPRA results and do not

appear to be fully resolved:

v. F&O 10-1 against SR HR-E1:

Step 11 of the procedure documented in Attachment F of the Post-Fire HRA report appears to indicate that a minimum value of 1.00E-07 was utilized for the joint probability of multiple HFEs. Section 6.2 of NUREG 1921 cites NUREG 1792 which advises that minimum joint HEPs "not be below ~1E-05 since it is typically hard to defend that other dependent failure modes that are not usually treated (e.g., random events such as even a heart attack)." If smaller than 1E-5 values were used, justify that in these cases that there is very low dependency between HFEs so that the acceptable minimum probability is not necessary supported by the results of a sensitivity study (i.e., CDF, LERF, CDF and LERF) that uses a 1E-5 minimum.

RESPONSE:

Using a 1E-5 floor for all HFE combinations does not give proper credit to long term decay heat removal (DHR) HFE dependencies. Combinations containing those long term DHR HFEs should have a 1E-6 floor for the following reasons. Long term DHR HFEs are those actions associated with using the containment vent to remove decay heat including post vent injection HFEs. Also included are HFEs associated with establishing late suppression pool cooling for DHR. These HFEs occur approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor scram, whereas most of the other HFEs occur within approximately six hours of the reactor scram. By the time these long term DHR actions are needed, a shift turnover would have occurred and an emergency response organization would have been implemented. A low dependency exists between these long term DHR HFEs and the earlier actions.

A sensitivity study compared the Fire PRA quantification metrics for the baseline case that utilized a floor of 1E-7 with a sensitivity Fire PRA quantification that utilized the above 1E-5 and 1E-6 floors. Tables PRA RAI 01.v-1 and PRA RAI 01.v-2 in Attachment 2 to this enclosure provide the results of the sensitivity and the values submitted in the LAR. Table PRA RAI 01.v-1 contains the results of the sensitivity study on the total fire CDF and LERF. Table PRA RAI 01.v-2 contains the results of the sensitivity study on the CDF and LERF. In subsequent Fire PRA quantifications, a joint probability floor of 1E-5 will be used for all combinations that do not include long term DHR HFEs. For combinations that include long term E1-19 DHR HFEs, a joint probability floor of 1E-6 will be used. After this change in method is implemented, the affected sections in the LAR will be revised to reflect results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to the remaining NRC

RAIs.

E1-20 PRA RAI 05

Section 2.4.3.3 of NFPA 805 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. In letter dated July 12, 2006, to NEI (ADAMS Accession No.

ML061660105), the NRC established the ongoing FAQ pr ocess where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. 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 discusses deviations and sensitivities from NUREG/CR-6850, but some apparent deviations may not have been addressed (e.g., related to F&O 2-54 and 2-56). Identify and describe all deviations from accepted methods and approaches, and clarify whether guidance from the June 21, 2012, memo from NRC to NEI, "Recent Fire PRA Methods review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires"' (ADAMS Accession No. ML 12171A583), was used in applying related methods. For identified deviations from NUREG/CR-6850 that fall outside this guidance memo, provide a sensitivity study that estimates the impact of their removal on CDF, LERF, CDF, and LERF. RESPONSE:

As noted in LAR Section 4.5.1.2, the process for creation of the Fire PRA model and quantification of that model used a methodology consistent with the guidance provided in NUREG/CR-6850 and subsequent clarifications documented in responses to NFPA 805 FAQs.

Deviations from NUREG/CR-6850 that were utilized in the Fire PRA model development are documented in LAR Attachment V.

The results of the sensitivity studies for these deviations are provided in LAR Attachment V.

These deviations and sensitivities, with the exception of the deviation discussed in LAR Section V.2.3, are not related to the guidance from the June 21, 2012, memo from NRC to NEI.

A discussion of the reduced transient heat release rates can be found in the TVA response to PRA RAI 16 (in the TVA letter dated February 13, 2014). The deviations and sensitivities discussed in LAR Attachment V are listed below and additional clarification is provided for any related changes.

1. LAR Section V.2.2 - "Generic Ignition Frequency Sensitivity Analysis" to support the use of updated generic fire frequencies arising from the industry review of fire events.

2. LAR Section V.2.3 - "Use of 69kW Heat Release Rate Used for Selected Transient Fires."

3. LAR Section V.2.4 - "Credit for very early warning fire detectors (VEWFDs) and Automatic Suppression for Fire Scenarios in Cable Spreading Room and Unit 1 Auxiliary Instrument Room." TVA has applied a prompt detection credit for area wide very early warning fire detection systems in the Unit 1 Auxiliary Instrument Room and the CSRs.

a. The area wide very early warning fire detection system is no longer being installed in the Auxiliary Instrument Rooms (See FPE RAI 10). A sensitivity study E1-21 was performed, as documented in the TVA response to FPE RAI 12, Part c (in this enclosure), to analyze the effect of removing the prompt detection credit. b. A sensitivity study was performed for the area wide very early warning fire detection system in the CSRs that is detailed in LAR Section V.2.4.

4. LAR Section V.2.5 - "Credit for Electrical Raceway Fire Barrier Systems (ERFBS) that are installed in accordance with NFPA 805 Chapter 3 Section 3.11.5." As discussed in the response to FPE RAI 5 (in the TVA letter dated January 10, 2014), this deviation has been removed. TVA has re-considered the use of Engineering Equivalency Evaluations for this application and decided not to disposition 1-hour ERFBS without automatic suppression as adequate for the hazard. The current plan for these applications is to install 1-hour ERFBS and to resolve the Variance from Deterministic Requirements using the fire risk evaluation process.

TVA has identified the following deviations that are not addressed in LAR Attachment V.

1. The TVA response to PRA RAI 01.f (in this enclosure) addresses the apparent deviations noted in the RAI for F&O 2-54. The TVA response to PRA RAI 01.f provides the results of a sensitivity study.

2. The catastrophic turbine/generator (T/G) fire postulated for PAU 26A is not consistent with Table O-2 of NUREG/CR-6850. The TVA response to PRA RAI 01.h, Part ii (in of this letter) provides the results of a sensitivity analysis.

3. The dependency analysis was not performed between non-LERF and LERF actions. The TVA response to PRA RAI 01.o (in this enclosure) states that in subsequent Fire PRA quantifications, the HRA will consider LERF HFE dependencies.

4. It is unclear whether those ignition sources that do not result in any scenarios that propagate to secondary targets were maintained in the analysis to reflect the fire-induced failure of the ignition source itself. The TVA response to PRA RAI 01.s (in TVA letter dated February 13, 2014) states that non-propagating scenarios that were originally screened have been identified and will be entered into SAFE-PB and quantified to include the risk contribution into the Fire PRA.

5. The joint probability floor value recommended in NUREG-1792 was not used in the Fire PRA. The TVA response to PRA RAI 01.v (in this enclosure) states that in subsequent Fire PRA quantifications, a joint probability floor of 1E-5 will be used for all combinations that do not include long term DHR HFEs. For combinations that include long term DHR HFEs, a joint probability floor of 1E-6 will be used. The TVA response to PRA RAI 01.v provides the results of a sensitivity analysis.

6. The MCR analysis indicates that frequency is apportioned by tray length and not weight or combustible loading as suggested by NUREG/CR-6850. The TVA response to PRA RAI 17.d (in TVA letter dated February 13, 2014) provides the basis for the approach used in the analysis.

The first five above described deviations will be included in the quantification in the next revision of the baseline Fire PRA model. After these changes in methods are implemented, the affected sections in the LAR will be revised to reflect the results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to remaining NRC RAIs.

E1-22 The apparent deviation noted for F&O 2-56 was addressed in the TVA response to PRA RAI 01.h, Part i (in TVA letter dated February 13, 2014), that provided the basis for the approach used in the analysis.

TVA used parts of the guidance described in the June 21, 2012, memo from NRC to NEI, "Recent Fire PRA Methods review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires,' " (hereinafter referred to as "the June 21 memo") as discussed below.

1. For "Frequencies for Cable Fires Initiated by Welding and Cutting," TVA incorporated the June 21 memo guidance for updated frequencies for cable fires initiated by welding and cutting, with no deviations. Therefore, a sensitivity study is not required.

2. Clarification for Transient Fires: For most fire compartments, TVA analyzed transient fires consistent with the guidance in NUREG/CR-6850, with no deviations. TVA applied the June 21 memo guidance to reduce the transient heat release rates in other fire compartments. This is discussed in LAR Section V.2.3 and the TVA response to PRA RAI 16 (in TVA letter dated February 13, 2014). Because TVA did not deviate from either of these NRC endorsed methods, a sensitivity study is not required.

3. Alignment Factor for Pump Oil Fires: With the exception of oil fire scenarios for Control Bay Chillers 3A and 3B in the CSR, TVA analyzed oil fires consistent with the guidance in NUREG/CR-6850, with no deviations. TVA's response to FM RAI 01.g (in TVA letter dated January 14, 2014) addressing oil fire scenarios for Control Bay Chillers 3A and 3B in the CSR included a sensitivity study using the oil alignment factors from the June 21 memo. Because TVA did not deviate from either of these NRC endorsed methods, a sensitivity study is not required.

4. Electrical Cabinet Fire Treatment Refinement Details: This guidance is not endorsed by the NRC and was not used by TVA.

5. EPRI 1022993, "Evaluation of Peak Heat Release Rates (HRRs) in Electrical Cabinet Fires": This guidance is not endorsed by the NRC and was not used by TVA.

TVA has identified no other deviations from accepted methods and approaches.

E1-23 PRA RAI 10

Section 2.4.3.3 of NFPA 805 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. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105, the NRC established the ongoi ng FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. 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.

Attachment V of the LAR indicates that area-wide incipient detection systems are credited in Fire Compartments 16-A, 16-K, 16-M and 16-O. For each applicable fire compartment,

a. Describe the credit given to incipient fire detectors in the fire PRA, including the resulting impact on fire scenario development (e.g., frequency apportionment, damage end states, timing, etc.).

b. Provide an assessment of the effectiveness of area-wide incipient detection systems and associated suppression system (as applicable) for fire scenarios analyzed and considering the characteristics of each compartment.

c. Justify the availability and reliability assumed for the area-wide incipient detector in all areas as well as the total flooding system planned for Fire Compartment 16-A.

RESPONSE:

As discussed in the TVA response to FPE RAI 10, TVA has determined that the proposed modification described in LAR Attachment S, Table S-2, Modification 77 to install area wide incipient (i.e. very early warning) detection systems, using very early warning fire detectors (VEWFDs), in the auxiliary instrument rooms (i.e., Fire Compartments 16-K, 16-M, and 16-O),

and discussed in LAR Attachment V will not be installed. LAR Attachment S, Table S-2, Modifications 78 and 79 propose modifications to install an area wide very early warning detection system and a new automatic gaseous fire suppression system in the cable spreading rooms (CSRs). Therefore, the remainder of this response addresses only the area wide systems in the CSR portion of Fire Compartment 16-A.

Part a As discussed in LAR Attachment V,Section V.2.4, the Fire PRA credits both prompt detection

and automatic suppression for transient fires, cable fires caused by welding and cutting, transient fires caused by welding and cutting, and self-ignited cable fires. This credit is based

on the VEWFDs (i.e., aspirating smoke detectors) to be installed as part of the very early warning fire detection system in the CSRs.

Current plans include VEWFDs to send pre-activation alerts to plant personnel, as discussed in the TVA response to FPE RAI 12, Part d. The detectors are not credited to detect the fire in its incipient phase. Instead, prompt detection and suppression is credited for preventing fire propagation beyond the fire origin only, which is within the guidance in NUREG/CR-6850, Appendix P for a high-sensitivity smoke detection system.

E1-24 Transient fires and self-ignited cable fires credited prompt detection, via the very early warning fire detection system, to actuate the automatic suppression system, which limited the target damage to one cable tray. When either prompt detection or automatic suppression was unsuccessful, all targets in the CSR were considered damaged.

Transient fires caused by welding and cutting and cable fires caused by welding and cutting credited prompt detection, via the very early warning fire detection system, to actuate the automatic suppression system and limit the fire damage to one cable tray. If the detection system was unsuccessful, the fire watch was credited to promptly detect the fire and manually suppress it, which also limited the target damage to one cable tray. If the detection system, automatic suppression system, or fire watch was unsuccessful, all targets in the CSR were assumed to be damaged.

Part b As discussed in the TVA response to FPE RAI 12, Part d, the CSR area wide detection system and the associated suppression system will be designed, installed, operated, and maintained to the specifications in NFPA 72, NFPA 76 for response transport times and sensitivity settings, NFPA 2001, "Standard on Clean Agents Fire Extinguishing Systems," and manufacturer listing requirements. Both systems will provide area wide coverage for the CSR. The system designs will incorporate adequate concentrations and the detectors and nozzles for the systems will be located such that a fire in any location of the CSR can be properly detected and/or extinguished

by these systems.

The existing guidance in NUREG/CR-6850, Appendi x P, recommends the treatment of high-sensitivity smoke detectors by crediting a prompt detection response. TVA recognizes that a very early warning fire detection system in an area wide application introduces an additional layer of complexity, in that the specific location of the fire may not be immediately apparent.

That being the case, it may not be until the fire has grown beyond the incipient phase (i.e., until it has entered the active burning/growth phase) that its location becomes apparent. Given this consideration, no credit is assumed in the Fire PRA for suppression to prevent damage to the source of the fire. Instead, prompt detection and suppression are credited for preventing fire propagation beyond the fire origin only.

Therefore, the area wide detection system and the associated suppression system will be able to effectively control the postulated fires in Fire Compartment 16-A, consistent with the guidance provided in NUREG/CR-6850, Sections 11.5.1.8.2 and P.1.2.

Part c The availability and reliability of the CSR area wide detection system is justified and explained in detail in the TVA response to FPE RAI 12, Part a.

A total flooding clean agent suppression system will be installed in the CSR portion of Fire Compartment 16-A. Applying an unavailability and unreliability value of 2E-02 for the area wide total flooding suppression system was appropriately included in the Fire PRA, based on the following:

The new total-flooding, clean agent system will be designed, installed, and maintained to the current NFPA standard (NFPA 2001) for Clean Agent Fire Extinguishing Systems.

NUREG/CR-6850 uses the failure probabilities (i.e., sy stem reliability) for automatic suppression systems based on NSAC-179L, "Automatic and Manual Suppression Reliability Data for Nuclear Power Plant Fire Risk Analyses," April 1994. NSAC-179L examined fire events up to 1988 that E1-25 are now over 26 years old. Many of these systems were likely installed more than 26 years ago. A new system will be designed and installed incorporating decades of technological improvements and code/standard refinement to significantly improve system reliability, for example: Later editions of NFPA 2001 contain more definitive requirements, calculation methods and guidelines for establishing minimum design concentrations, safety factors, agent quantity, hold times, etc., for all types of fires, including deep-seated fires. Additional design factors include: unclosable openings, acid gas formation, fuel geometry, enclosure geometry, obstructions, uncertainty of agent quantity distribution at tee splits, and agent lost through initial compartment relief venting. Enhanced electronic supervision capabilities for all system components and features necessary for proper system operation. Current industry standard practice includes audible and visual indication of all adverse conditions (e.g., low cylinder pressure, wiring/circuit faults, damper status/closure, actuating devices, power interruption).

The fire scenarios crediting the total flooding suppression system in the Fire PRA account for the unreliability and unavailability of the detection system with a value of 1E-02, as discussed in the TVA response to FPE RAI 11, Part b, independently. This is conservative when compared to the non-suppression probability values in NUREG-6850, Appendix P, for gaseous suppression systems, which include detection failures.

In addition, the sensitivity analysis discussed in LAR Section V.2.4 analyzed the effect of changing the total probability of non-suppression to 5E-02. Selecting a non-suppression probability of 5E-02 bounds the reliability of Halon, deluge and pre-action systems, which have the lowest reliability of the recommended values in NUREG/CR-6850, Appendix P. As a result of this sensitivity, BFN meets the guidance for a Region II plant with total CDF and LERF below 1E-04/rx-yr and 1E-05/rx-yr, respectively, for ov erall plant risk. BFN also meets the CDF/LERF criteria for a Region II plant which allows a positive CDF of 1E-05/rx-yr and LERF of 1E-06/rx-yr for acceptable risk increases.

E1-26 PRA RAI 12

Section 2.4.3.3 of NFPA 805 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 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.

Item 3 under Section 4.2.1 of the Post-Fire HRA report indicates that some modeled actions were divided into short- and long-term actions to account for system failures; however, the criteria for which action is credited are noted as being based on fire-induced failures alone. Given this non-conservative treatment, provide the results of a sensitivity study (i.e., CDF, LERF, CDF and LERF) that applies short- and long-term actions considering the impact of system failures due to both fire-induced and random equipment failures.

RESPONSE:

The Fire PRA quantified the same HFE differently for different fire scenarios, depending upon the equipment that is failed by the fire in a given scenario and the operator response timing differences for each scenario. Because timing is a key input to human error probability, a longer time frame for operator response provided the opportunity for additional resources (e.g., Shift Technical Advisor, Emergency Response Facility) and greater likelihood for correct diagnosis and execution performance. In addition, using different HFEs provided a more accurate representation of the situation presented to the operator in terms of the plant functional state and equipment available for use in operator response.

The statement about short-term and long-term actions in Item 3 under Section 4.2.1 of the Post-Fire HRA report applies to the emergency depressurization actions discussed under Section 5.1.6 of the same report. Two HFEs were developed to model emergency depressurization actions in two time frames (i.e., 30 minutes and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />). Neither of these HFEs considered random failures of high pressure injection to determine which timing would be used; only fire effects were considered, which is non-conservative. The model also included restricted depressurization timing based on fire compartment effects (e.g., if high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) are failed at the fire compartment level all scenarios in the compartment assumed early depressurization would be required), which is a conservative assumption.

A sensitivity study was performed in which both assumptions were removed from the models. In this study, depressurization timing depends on both fire-induced and random failures and is scenario specific. Tables PRA RAI 12-1 and PRA RAI 12-2 in Attachment 2 to this enclosure provide the results of the sensitivity study and a comparison to the values submitted in the LAR.

Table PRA RAI 12-1 contains the results of the sensitivity study on the total fire CDF and LERF.

Table PRA RAI 12-2 contains the results of the sensitivity study on the CDF and LERF. The Fire PRA model will be revised to consider random failures of high pressure injection when determining the timing for emergency depressurization. After this change in method is implemented, the affected sections in the LAR will be revised to reflect results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to remaining NRC RAIs.

E1-27 PRA RAI 15

Section 2.4.3.3 of NFPA 805 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. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoi ng FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. 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 justif ication to allow the NRC staff to complete its review of the proposed method.

Attachment 8 of the Fire-Induced Risk Model report describes treatment for shorting switches where a failure probability of 1E-3 (grounded circuits) is applied. This approach is stated to be reasonably bounding; however, no basis is provided for these values to assess their level of conservatism. Provide justification for the applied probabilities, or provide the results of a sensitivity study (i.e., CDF, LERF, CDF and LERF) that assumes failure of shorting switches under conditions involving cables of interest that are co-located with higher voltage level cables or cables with potentially higher fault current.

RESPONSE:

The justification for crediting the shorting switches is based on the discussion in TVA Fire PRA - Task 7.5 Fire Induced Risk Model, Attachment 8. Credit is taken in the Fire PRA model for shorting switches associated with the following valves:

Component Description Type Circuit Existing Design/Modification Number (from LAR Table S-2) 1-FCV-074-0052 Division (Div) 1 Low Pressure Coolant Injection (LPCI)

Injection Valve Motor Operated Valve (MOV) Ungrounded

Control Power Transformer (CPT) Existing Design 1-FCV-074-0066 Div 2 LPCI Injection Valve MOV Ungrounded CPT Existing Design 2-FCV-074-0052 Div 1 LPCI Injection Valve MOV Ungrounded CPT Existing Design 2-FCV-074-0066 Div 2 LPCI Injection Valve MOV Ungrounded CPT Existing Design 3-FCV-074-0066 DIV 2 LPCI Injection Valve MOV Ungrounded CPT Existing Design 1-PCV-069-0015 Reactor Water Cleanup (RWCU) Blowdown Valve Air Operated Valve (AOV) 4-20 milliamps (mA) current

loop Modification number

90 2-PCV-069-0015 RWCU Blowdown Valve AOV 4-20 mA current loop Modification number

90 E1-28 Component Description Type Circuit Existing Design/Modification Number (from LAR Table S-2) 3-PCV-069-0015 RWCU Blowdown Valve AOV 4-20 mA current loop Modification number 90 Credit is taken in one of three ways depending on the configuration and the location of the fire as follows:

Case 1 The control circuits were affected by fire, but the shorting switch, shorting switch conductors, and actuated device remained free of fire damage. The "line side" of the actuation device of concern maintained electrical continuity to the circuit return path via the shorting switch, thus allowing the design function of the shorting switch to remain intact. Consequently, no credible hot shorts to other conductors also connected to the "line side" of the actuation device result in energization of the actuation device because the source conductor is immediately shorted to the circuit return path.

For the MOV control cables, one potential uncertainty discussed in TVA Fire PRA - Task 7.5 Fire Induced Risk Model, Attachment 8 is the effect that an inter-cable hot short from a higher voltage (i.e., 480 Volts (V) and above) or a higher current carrying capacity circuit cable may have on the shorting switch function. The Fire PRA considers this to be unlikely. BFN does not generally route control cables and higher voltage cables in the same raceway. However, if they are routed in the same raceway, they are separated by a metal barrier which reduces the likelihood that an inter-cable hot short between power and control cable applying the high voltage condition to the shorting switch circuit would occur. Although the metal barrier can be considered to prevent any inter-cable interactions between the high and low voltage cables, potential failure modes of this configuration were considered. Any inter-cable hot short would need to occur after fire-induced melting or destruction of the cable tray barrier concurrent with another inter-cable hot short on the circuit return path, or the control cable would need to short circuit to the cable tray ground plane concurrent with a short of the higher voltage cable to the ground plane and an inter-cable hot short of a different phase of the higher voltage cable on the circuit return path. These combinations of concurrent failures were considered incredible and spurious operation was excluded for this case.

The AOVs are controlled by an instrument circuit and positioner. The shorting switch will disconnect the controller from the positioner and short the conductors to the positioner. In order for a spurious operation to occur, there must be an open circuit on the shorting switch (that is free of fire damage for Case 1), then a short circuit that applies a proper amplitude signal voltage to the positive signal cable concurrent with a short circuit of the aggressor return path to the negative signal cable. These combinations of concurrent failures were considered incredible and spurious operation was excluded for this case.

Case 2 For MOV control circuits, spurious operation was considered possible if the field cables containing the shorting conductors were exposed to fire damage. Fire exposure was not postulated to cause open circuits as the primary failure mode. Typically, insulation would have been fully consumed before the copper conductor melted and potential target and source conductors would have already grounded or thei r fuse/breaker would have cleared. One E1-29 potential failure mode was electrical arcing from high energy alternating current (AC) or direct current (DC) circuits causing collateral damage to the shorting switch conduction path, resulting in an open circuit of the shorting switch conductors and failure of the shorting switch function. BFN raceways are allowed to contain cables for AC and DC circuits, so a probability of 1E-03 was used to model the likelihood that a cable containing shorting switch conductors was located next to high energy circuit cable capable of generating an arc that caused collateral damage and potential open circuit of the shorting switch conductors. Although there is uncertainty associated with this probability, it bounds the likelihood that the cable containing shorting switch conductors was located next to a cable capable of generating the arc, that the arc was generated when the cable was affected by a fire and results in an open circuit of the shorting conductors, and that after the open circuit occurs, a fire-induced short causes a spurious operation.

The effect of a potential inter-cable hot short from a higher voltage exists as described in Case 1. The treatment given to cables containing conductors associated with the shorting switch is the same as that described for Case 1.

For the AOV positioner signal cables, the shorting switch design maintains the conductors in a shorted position, so the same consideration applies and 1E-03 is used to model the likelihood that both an instrument cable was located next to a high energy circuit cable capable of generating an arc that caused collateral damage and potential open circuit of the cable conductors and a subsequent hot short of the correct polarity and amplitude occurred through the grounded shield.

Case 3 The shorting switch and associated panel wiring were exposed to fire damage. All of the shorting switches for both the MOV and the AOV control circuits are located or, in the case of the modifications, will be located in the Main Control Room panels in Fire Compartment 16-A. In the Fire PRA, a probability of 1E-03 was used to bound the likelihood of a fire induced hot short concurrent with an open circuit of the shorting switch conduction path that fails the shorting switch function; however, estimation of the failure probability of the shorting switch in these conditions is beyond the current state of knowledge and little test data exists. Similar to Case 2 for the MOV circuits, insulation on the shorting switch conductor would have to be fully consumed before the copper conductor melts, or the terminal strips, shorting switch, or other devices in the shorting conduction path would have to be consumed/destroyed prior to potential target and source conductors shorting together.

For the AOV circuits, the 1E-03 probability is bounding because both the shorting switch conduction path must fail open followed by a hot short on the positive signal cable with the correct voltage signal amplitude and a short circuit occurs on the return path on the negative signal cable as described in Case 1.

E1-30 PRA RAI 20

Section 2.4.3.3 of NFPA 805 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 NRC.

RG 1.174 provides quantitative guidelines on CDF, LERF, 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.

The post transition (or variant) plant will continue to use a number of existing, non-fire specific OMAs (e.g., actions to align and start the swing emergency equipment cooling water train and emergency depressurization actions). During the audit, TVA clarified that some of these existing OMAs, whose completion is unaffected by fire, are not credited (i.e., set to failed) when quantifying the risk of the compliant plant. Address the following:

a. Clarify whether any of the non-fire specific OMAs are credited in the compliant plant risk calculations.

b. Summarize all non-fire specific OMAs that are unaffected by fire but that are set to failed while quantifying the compliant plant risk.

c. Provide the criteria for deciding which existing, non-fire specific OMAs were credited and which ones were set to fail when quantifying the compliant plant risk.

d. Explain how setting OMAs unaffected by fire to failed is consistent with the guidance that the risk estimate is a realistic estimate of the compliant plant risk as it would be operated after each fire.

RESPONSE:

Not all non-fire specific operator actions that occur outside of the MCR were credited for each Fire Area in the calculations supporting the LAR. To summarize, these actions are operator actions occurring outside the control room that are modeled in the Internal Events PRA models and are proceduralized in either the Operating Instructions, Abnormal Operating Instructions (AOIs), or Emergency Operating Instructions (EOIs). In the calculations supporting the LAR, each non-fire specific operator action occurring outside the control room modeled in the Fire PRA was selectively credited on a Fire Area basis where needed to support risk reduction. Credit was given in both the post-transition and the compliant plant risk models until a level of acceptable risk was reached in the analysis.

In lieu of providing justification in response to PRA RAI 20, Part d, the post-transition and compliant plant Fire PRA models will be revised to include non-fire specific, operator actions occurring outside the control room to more realistically model the fire risk, as opposed to only those that were selected for initial risk reduction. Inclusion of these non-fire specific, operator actions occurring outside the control room will result in a decrease in both the post-transition and the compliant plant total per unit fire risk. A sensitivity study was performed that estimates the risk decrease attributed to the inclusion of the non-fire specific, operator actions occurring outside the control room. Tables PRA RAI 20-1 and PRA RAI 20-2, in Attachment 2 to this enclosure, provide the results of the sensitivity study and a comparison to the values submitted in the LAR. Table PRA RAI 20-1 contains the results of the sensitivity study on the total fire E1-31 CDF and LERF. Table PRA RAI 20-2 contains the results of the sensitivity study on the CDF and LERF. After this change in method is implemented, the affected sections in the LAR will be revised to reflect the results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to remaining NRC RAIs.

E1-32 PRA RAI 23.d

Section 2.4.3.3 of NFPA 805 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 ASME/ANS 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 primary results of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os.

Clarify the following dispositions to fire F&Os and SR assessment identified in Attachment U of the LAR that have the potential to impact the fire PRA results and do not appear to be fully resolved:

d. 2-41 (Finding against LE-C6 Not Met at CC-I/II/III)

The disposition to the F&O appears to indicate that plant system models were not developed before the peer review. Describe what models were developed and how they have been reviewed.

RESPONSE:

The Internal Events Peer Review Team reviewed a draft version of the LERF Notebook. The LERF Notebook, Revision 0, which describes and documents the methodology used to determine relevant LERF functions, was issued subsequent to the peer review. LERF sequences were identified through the development of a series of containment event trees (CETs). Separate CETs were developed for each core damage functional class that could result in large early releases. The calculation documenting this treatment has not yet been subjected to peer review with respect to this supporting requirement.

The LERF analysis interfaces with the Level 1 accident sequence analysis through the appropriate set of core damage functional classes. A set of functional fault trees were used to describe the failure modes at each event tree node. The Level 1 sequences endpoints became the initiator for the CETs.

Active Containment and Reactor Building systems analysis important to a Level 2 analysis were developed including isolation, drywell spray and st andby gas treatment. The active containment systems were included in the Level 1 model such that the resulting CETs examine phenomenology and associated uncertainties. That examination of augmenting CETs with functional fault trees (i.e., system modeling) provides the basis for the probabilities assigned to the CETs.

LAR Attachment U, Table U-1, Supporting Requirement LE-C6 (i.e., F&O 2-41), "Resolution Section" is revised to state: "Active containment and reactor building system analysis important to Level 2 were developed in the Level 1 model including isolation, drywell spray and standby gas treatment. Containment event trees (CETs) were augmented with functional fault trees (system modeling) that provided the basis for probabilities assigned to the CETs."

As described in the TVA response to PRA RAI 11.b and 11.c (in the TVA letter dated January 14, 2014), a focused scope peer review will be performed prior to transition to NFPA 805. To ensure this peer review is performed, the TVA response to PRA RAI 11.b E1-33 and 11.c added a new implementation item, Implementation Item 47, to LAR Table S-3. The LAR Table requires that certain items be completed prior to the implementation of the NFPA 805 fire protection program. This response to PRA RAI 23.d revises the proposed Implementation Item 47 provided in the response to PRA RAI 11.b and c to include the LERF Analysis (element LE-C6) as follows: "Perform a focused-scope peer review of the Fire PRA.

The peer review will include, as a minimum, the following elements: Fire PRA Cable Selection and Location (CS), Human Reliability Analysis (HRA), Fire Risk Quantification (FQ), Uncertainty and Sensitivity Analysis (UNC), and LERF Analysis (element LE-C6)."

E1-34 RR RAI 01

For several compartments described in LAR Attachment E, there are potential liquid and gaseous effluents discharged from the plant during fire suppression activities.

a. Provide a general description of the storm drain system outside the plant and the amount of liquid effluent dilution that would occur prior to reaching a member of the public.

b. Provide a general description of the distance to the site boundary and the amount of gaseous effluent dispersion that would occur prior to reaching to members of the public.

RESPONSE:

Part a

The storm drain system collects stormwater runoff from areas, components and structures throughout the facility. The storm drain system is comprised of yard drains with a subsurface piping network, open channels and ditches. These systems discharge to Wheeler Reservoir either directly or by way of the intake forebay or yard drainage basin. The storm drains do not contain any immediate method of isolating or containing flow without manual means.

Part b With respect to the distance to the site boundary, the building containing radioactive material whose location has the shortest potential distance to members of the public is the Low Level Radwaste Tool Warehouse. This building is located in the owner controlled area, approximately one-eighth mile from a public road.

Parts a and b

In lieu of providing the amount of liquid effluent dilution or gaseous effluent dispersion that would occur prior to reaching members of the public, the following discussion is provided. LAR Section 4.4 describes the qualitative defense in depth process to evaluate the potential for radioactive release due to firefighting activities. This qualitative defense in depth process includes:

Determining the installed engineering controls for the compartments that have the potential to contain radioactive materials. LAR Attachment S, Table S-3, Implementation Item 1.g requires the development of administrative controls to support actions to prevent radioactive release. The intent of the administrative controls is to require site procedures to provide options for compliance in these areas. For example, materials may be stored in metal containers with tight fitting closures and/or covers. This would contain the radioactive material during fire suppression activities. Where it is not practical to store radioactive materials in tight fitting metal containers, a source term evaluation will be completed to establish appropriate administrative controls to ensure that a fire involving radioactive material will not exceed 10 CFR Part 20 limits. This evaluation will consider various input parameters such as type and quantity of fire loading, type of firefighting suppression, levels of loose (dispersible) radioactive contamination available for release, and the effluent dispersion and dilution factors as needed based on the specific configuration of the analyzed areas. LAR Attachment S, Table S-3, Implementation Item 1.h requires training each fire brigade member to identify potential points for radioactive release and the actions that E1-35 can be taken to mitigate a release. In addition, Attachment S, Table S-3, Implementation Items 1.a and 1.e require guidance outlining these expectations and actions to be provided in pre-fire plans and standard operating procedures.

The use a combination of qualitative and quantitative analyses (as needed) to minimize radioactive releases and the defense in depth measures will ensure that radioactive releases are below 10 CFR Part 20 limits during fire suppression activities.

E1-36 RR RAI 02 a) In the Service Building, describe whether there any ventilation controls or floor drains. If yes, describe those engineered systems.

b) For the East Access Building, there was no credit taken for specific ventilation controls or floor drains. Instead, the installed sprinkler system, administrative controls and standard operating procedures are credited to contain both gaseous and liquid releases.

c) For the South Access Building and Outage Rad Materials Warehouse, there were no specific ventilation controls or floor drains credited.

d) For the Standby Gas Treatment Building, there were no specific ventilation controls or floor drains credited due to the remote location of the facility and the potential for engineering controls to be affected by a fire.

e) For the Off-Gas Building and Off-Gas Stack, there were no specific ventilation controls or floor drains credited due to the louvered air intakes and door openings, and the remote location of the facility and the potential for engineering controls to be affected by a fire.

f) For the YARD-Radiological Controlled Area, the yard area is open to atmosphere and has no mechanisms to assist or act as an engineering control, and yard storm drains collect water runoff.

g) For the Low Level Radwaste Tool Warehouse, there was no credit taken for specific ventilation controls or floor drains. Instead, the installed sprinkler system and administrative controls are credited to contain a gaseous effluent releases.

Administrative controls and standard operating procedures are credited to contain both gaseous and liquid releases. For each building/area described above, provide a description of the administrative controls and standard operating procedures and describe how effective they will be in providing containment of gaseous and liquid releases. If adequate containment cannot be assured, describe the input parameters (e.g., fire loading, type of firefighting suppression, source term; i.e., level of loose (dispersible) radioactive contamination available for release, and the effluent dispersion and dilution factors) that would be used in a qualitative or quantitative evaluation that describes compliance with the limits described in 10 CFR Part 20.

RESPONSE:

Each of the plant areas listed in RR RAI 02 has limited or no engineering controls with respect to ventilation and drainage. The evaluation conducted to provide input to LAR Attachment E is qualitative. In the locations listed in the RAI, engineering controls could not be credited to support radioactive release goals. The engineered systems and features that are present are identified in LAR Attachment E. Because these systems and/or features cannot contain the possible airborne and/or liquid effluent during fire suppression activities, administrative controls are the credited method for compliance. Each area is addressed as follows:

a) The Service Building does not have ventilation controls or floor drains that route to a monitored release point. Administrative controls are the only credited method of compliance.

E1-37 b) The East Access Building does not have ventilation controls or floor drains that route to a monitored release point. The reference to the installed sprinkler system in the building is intended to identify that the feature exists and can provide benefit when available. Administrative controls are the only credited method of compliance.

c) The South Access Building and Outage Rad Materials Warehouse do not have ventilation controls or floor drains that route to a monitored release point. Administrative controls are the only credited method of compliance.

d) The Standby Gas Treatment Building has engineering controls in the form of a building ventilation system and floor drains, but is still susceptible to releasing smoke and/or fire suppression runoff. Administrative controls are the only credited method of compliance.

e) The Off-Gas Building and Off-Gas Stack have engineering controls in the form of a building ventilation system and floor drains, but are still susceptible to releasing smoke and/or fire suppression runoff. Administrative controls are the only credited method of compliance.

f) The Yard-Radiological Controlled Area has no engineering controls. Administrative controls are the only credited method of compliance.

g) The Low Level Radwaste Tool Warehouse does not have ventilation controls or floor drains that route to a monitored release point. The reference to the installed sprinkler system in the building is intended to identify that the feature exists and can provide benefit when available. Administrative controls are the only credited method of compliance.

The following discussion describes the defense in depth process associated with containment of gaseous and liquid releases in the buildings/areas described in the RAI. LAR Section 4.4 describes the qualitative defense in depth process to evaluate the potential for radioactive release due to firefighting activities. This qualitative defense in depth process includes:

Determining the installed engineering controls for the compartments that have the potential to contain radioactive materials. LAR Attachment S, Table S-3, Implementation Item 1.g requires the development of administrative controls to support actions to prevent radioactive release. The intent of the administrative controls is to require site procedures to provide options for compliance in these areas. For example, materials may be stored in metal containers with tight fitting closures and/or covers. This would contain the radioactive material during fire suppression activities. Where it is not practical to store radioactive materials in tight fitting metal containers, a source term evaluation will be completed to establish appropriate administrative controls to ensure that a fire involving radioactive material will not exceed 10 CFR Part 20 limits. This evaluation will consider various input parameters such as type and quantity of fire loading, type of firefighting suppression, levels of loose (dispersible) radioactive contamination available for release, and the effluent dispersion and dilution factors as needed based on the specific configuration of the analyzed areas. LAR Attachment S, Table S-3, Implementation Item 1.h requires training each fire brigade member to identify potential points for radioactive release and the actions that can be taken to mitigate a release. In addition, Attachment S, Table S-3, Implementation Items 1.a and 1.e require guidance outlining these expectations and actions to be provided in pre-fire plans and standard operating procedures.

E1-38 The use a combination of qualitative and quantitative analyses (as needed) to minimize radioactive releases and the defense in depth measures will ensure that radioactive releases are below 10 CFR Part 20 limits during fire suppression activities.

ATTACHMENT 1 TO ENCLOSURE 1 Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 FPE RAI 10 - LAR markup pages TVA BFN Attachment V - Fire PRA Quality Page V-6 1E-05/rx-yr for overall plant risk assuming either 10% or 67% of the transient fire frequency results in a full compartment burn (FCB). Browns Ferry Nuclear meets the CDF/LERF criteria for a Region II plant which allows a positive delta CDF of 1E-05/rx-yr and LERF of 1E-06/rx-yr for acceptable risk increases for both cases in Tables 4 and 6 of the calculation. V.2.4 Credit for VEWFDs and Automatic Suppression for Fire Scenarios in Cable Spreading Room and Unit 1 Auxiliary Instrument Room The BFN Fire PRA applies a prompt detection credit and an automatic suppression credit for transient fires, cable fires caused by welding and cutting, transient fires caused by welding and cutting, and self-ignited cable fires in the Cable Spreading Rooms (Fire Area 16/Fire Compartment 16-A). This is based on the planned modification to install aspirating smoke detectors (ASD) installed as very early warning fire detectors (VEWFDs) that will actuate a total flooding clean agent suppression system (Refer to Attachment S, Table S-2, items 78 and 79 for additional details on the modifications). The detectors are credited with prompt detection, not detection of the fire in its incipient phase.The credit applied for the planned VEWFDs as a prompt detector is considered

acceptable since: The credit for prompt detection is consistent with the guidance contained in NUREG/CR-6850, Appendix P, for crediting High sensitivity detectors, and specifically aligns with the Detection-Suppression Event Tree Output for sequences A though E, as contained in Table P-1; The detection system availability and reliability of 1E-02 from FAQ 08-0046 is appropriately included; The appropriate manual fire suppression failure probabilities in NUREG/CR-6850, Supplement 1 (i.e., FAQ 08-0050) are selected based on the specific initiator; Even when the VEWFDs and manual suppression are successful, the tray damaged by the fire is conservatively assumed to be one of the top 25 risk contributing cable trays in Fire Compartment 16-A, for each unit. Transient scenarios have been quantified for the 25 top risk significant cable trays in each unit. A weighting factor for these scenarios was calculated and applied based on the length of each cable tray divided by the total length of the trays; When either the VEWFDs or manual suppression are unsuccessful, the analysis conservatively assumes whole room damage; and The majority of the cable trays in the cable spreading room are provided with Flamemastic coating and/or bottom covers which would delay damage and ignition, however, no credit was given for this in the analysis.

The credit for the planned total flooding clean agent suppression system is considered acceptable since: The credit for automatic suppression is consistent with the guidance contained in NUREG/CR-6850, Appendix P, for crediting High sensitivity detectors together with automatic suppression, and specifically aligns with the Detection-Suppression Event Tree Output for sequence B, contained in Table P-1; The automatic suppression availability and reliability of 2E-02 is appropriately included; Even when the VEWFDs and automatic suppression system are successful, target damage is conservatively assumed to be one of the top 25 risk contributing cable trays

in Fire Compartment 16-A, for each unit, as described above; BFN Units 1, 2, and 3 NFPA 805 Transition Report, Page 1446 of 1661 TVA BFN Attachment V - Fire PRA Quality Page V-7 When either the VEWFDs or both automatic and manual suppression system are unsuccessful, the analysis conservatively assumes whole room damage; and The majority of the cable trays in the cable spreading room are provided with Flamemastic coating and/or bottom covers which would delay damage and ignition, however, no credit was given for this in the analysis. The BFN Fire PRA also applies a prompt detection credit for two transient fire scenarios in the Unit 1 Auxiliary Instrument Room (Fire Area 16/Fire Compartment 16-K). The BFN Fire PRA does not apply a prompt detection credit for any fire scenarios in the Unit 2 and 3 Auxiliary

Instrument Rooms (Fire Area 16/Fire Compartments 16-M and 16-N). The credit applied for the planned VEWFDs as a prompt detector is considered acceptable for the same reasons described above for the Cable Spreading Room.

Based on the discussion above, the approach implemented is considered conservative, and in majority aligns with guidance within NUREG/CR-6850, Appendix P. Therefore, a sensitivity

study on crediting VEWFDs together with a total flooding clean agent suppression system is not required. However, a sensitivity has been performed in TVA Calculation NDN0009992013000132, BFN Fire PRA - NFPA 805 Application Calculation, Revision 1, to illustrate that an increased probability of inadvertent lockout of the fire suppression system has

only a small effect on the results of the PRA. The sensitivity to model an increased probability of inadvertent lockout of the fire suppression system in the Cable Spreading Room was performed by increasing the non suppression probability for Transient Fires, Transient Fires due to Welding and Cutting, Self-Ignited Cable and Junction Box Fires, Cable Fires due to Welding and Cutting to 5.0E-02 for all four fire scenarios. The use of a 5.0E-02 as a probability of non-suppression for sensitivity purposes is

based on that of Halon, Deluge or Pre-Action, which have the lowest reliability of the recommended values in NUREG/CR-6850, Appendix P (Section P.1.3 on Page P-6) for automatic suppression systems. A sensitivity for the credit for prompt detection in the two Unit 1 Auxiliary Instrument Room transient fire scenarios was also performed by removing the credit for the VEWFDs altogether. The results of these sensitivity studies can be seen in Tables 7 and 8 of TVA Calculation NDN0009992013000132, Revision 1. As can be seen in Table 7 of the calculation, BFN meets the guidance for a Region II plant with total CDF and LERF below 1E-04/rx-yr and 1E-05/rx-yr for overall plant risk with the above sensitivities. BFN also meets the CDF/LERF criteria for a Region II plant which allows a positive delta CDF of 1E-05/rx-yr and LERF of 1E-06/rx-yr for acceptable risk increases with the above sensitivities in Table 8 of the calculation. V.2.5 Credit for Electrical Raceway Fire Barrier Systems that are installed in accordance with NFPA 805 Chapter 3 Section 3.11.5 Browns Ferry Nuclear has Electrical Raceway Fire Barrier Systems (ERFBS) installed, and modifications planned to install ERFBS, in accordance with NFPA 805 Chapter 3 Section 3.11.5 to meet the separation requirements of NFPA 805 Chapter 4 Section 4.2.4. In most cases the ERFBS are 1-hour rated and are or will be installed in areas where automatic detection and suppression is available. In some cases, the 1-hour rated ERFBS may not have automatic suppression. In those instances, an Engineering Equivalency Evaluation will be performed to determine that the 1-hour rated ERFBS is adequate for the hazard. The ERFBS are referenced in Tables C-1 and C-2 as a required Fire Protection System/Feature for the Fire Area. In addition, the engineering equivalency evaluation is referenced in Tables C-1 and C-2. In order to take credit for the ERFBS in the Fire PRA, two fire scenarios were developed for ignition sources that would damage the PRA targets that have ERFBS. One scenario credits BFN Units 1, 2, and 3 NFPA 805 Transition Report, Page 1447 of 1661 E2-1 ENCLOSURE 2 Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 Updated TVA Response to NRC Request for Additional Information FM 03, Part b.i and PRA 01.h, Part ii FM 03, Part b.i TVA previously responded to FM RAI 03, Part b.i, in a letter dated January 10, 2014 (CNL 14-001). Subsequent to the submittal of the response, TVA noted an error in the reference cited for validation of the Beyler method. Specifically, the reference was cited as MDQ0009992012000099, "Verification and Validation of Fire Modeling Tools and Approaches for use in NFPA 805 and Fire PRA Applications," Appendix A when it should have been MDQ0009992012000099, "Verification and Validation of Fire Modeling Tools and Approaches for use in NFPA 805 and Fire PRA Applications," Revision 2, Sections E.1 and E.7. The below response supersedes the previous response for FM RAI 03, Part b.i. The change from the previous response is shown with deleted text struck through, inserted text in bold, underline, and a revision bar to the right.

RESPONSE Beyler's method to estimate the HGL [hot gas layer] temperature in closed compartments was used in the MCA [Multi-Compartment Analysis] in accordance with NUREG-1805 Fire Dynamics Tool (FDT) 2.3, "Predicting Hot Gas Layer Temperature in a Room Fire with Door Closed." This FDT is discussed in NUREG-1805, Section 2.6 and verified and validated in NUREG-1824, ["Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications,"]

Volume 3, Section 3.1.2. The correlation was applied using verified methods within the validation range and limitations, or the use was justified as acceptable, as further discussed in RAI responses FM 03.b.ii and FM 03.b.iii.

The following line item should be added into Attachment J of the LAR:

E2-2 Calculation Application V&V Basis Discussion Hot Gas Layer (Method of Beyler) Calculates the

hot gas layer

temperature

for a closed compartment with no ventilation. NUREG-1805, Chapter 2, 2004 NUREG-1824, Volume 3, 2007 SFPE Handbook, 4 th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008 MDQ0009992012000099, "Verification and Validation of Fire Modeling Tools and

Approaches for use in NFPA 805 and Fire PRA Applications," Appendix A Revision 2, Sections E.1 and E.7 The correlation is used in the NUREG-1805 fire

model, for which V&V

was documented in NUREG-1824. The correlation is documented in an authoritative publication

of the SFPE Handbook of Fire Protection Engineering. The correlation has been applied within the

validated range

reported in NUREG-1824 or, if applied outside the validated range, the

model has been justified as acceptable, either by qualitative analysis, or by quantitative sensitivity analysis.

PRA 01.h, Part ii TVA previously responded to PRA RAI 01.h in a letter dated February 13, 2014 (CNL 14-020).

Subsequent to the submittal of the response, TVA noted a discrepancy in the delta risk calculation. Specifically, in the second table in the Part ii response the Fire Areas that were deterministically compliant were not set to a delta risk of N/A (i.e., 0.00) as was done in the BFN NFPA 805 LAR. The below response to Part ii supersedes the previous response for PRA RAI 01.h, Part ii. The change from the previous response is shown with deleted text struck through, inserted text in bold, underline, and a revision bar to the right.

RESPONSE NUREG/CR-6850, Table O-2 lists the severity factor values for catastrophic turbine generator fires. These are the fires postulated to generate damage to structural steel elements generating structural collapse. As requested in the RAI, the ignition frequency for the catastrophic scenarios has been updated to be the sum of bins 33, 34, and 35 (i.e., 2.10E-03+

3.23E-03+3.89E-03 = 9.22E-03). The severity factor of 0.025 was then applied with this E2-3 frequency as NUREG/CR-6850, Table O-2 recommends for "T/G fires involving H 2, oil, and possibly blade ejection." The risk parameters used to calculate the ignition frequency for the scenarios in the three BFN units are:

Unit Scenario Ignition Frequency (IGF) Severity Factor Non-Suppression Probability (NSP) 1 26-A.146-TGEX-CAT 9.22E-03 2.50E-02 1.10E-01 2 26-A.490-TGEX-CAT 9.22E-03 2.50E-02 1.10E-01 3 26-A.757-TGEX-CAT 9.22E-03 2.50E-02 1.10E-01 The Fire PRA model and documentation will be updated to reflect the values in the above table.

The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional

quantification is performed in response to all the NRC RAIs.

The sensitivity analysis results for Fire CDF and LERF are:

Unit Sensitivity (Plant Fire CDF) Percent (%) Increase From Baseline Sensitivity (Plant Fire LERF) % Increase From Baseline Sensitivity Delta Risk (COMP) CDF % Increase From Baseline Sensitivity Delta Risk (COMP) LERF % Increase From Baseline 1 6.29E-05 0.03% 2.14E-06 0.00% -5.59 37E-04 0.00% 1.91 93E-07 0.00% 2 6.59E-05 0.03% 1.90E-06 0.00% -4.89 66E-04 0.00% 3.10 20E-08 0.00% 3 5.30E-05 0.09% 1.83E-06 0.00% -5.68 45E-04 0.00% 1.18 19E-07 0.00%

In summary, the ignition frequency for the catastrophic turbine generator fires was updated to account for the sum of the generic frequencies in bins 33, 34, and 35. A sensitivity analysis was conducted to assess the effect on plant Fire CDF, LERF, CDF, and LERF for each of the three units. The effect in the plant fire risk value is minimal.

E3-1 ENCLOSURE 3 Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 Summary of BFN NFPA 805 RAI Response Dates RAI Question Number Type of Response (days) Actual Date of Response Fire Protection Engineering (FPE)

FPE 01 60 CNL-13-141 December 20, 2013 FPE 02 60 CNL-13-141 December 20, 2013 FPE 03 60 CNL-14-001 January 10, 2014 FPE 04 60 CNL-14-001 January 10, 2014 FPE 05 60 CNL-14-001 January 10, 2014 FPE 06 60 CNL-13-141 December 20, 2013 FPE 07 60 CNL-13-141 December 20, 2013 FPE 08 60 CNL-14-006 January 14, 2014 FPE 09 60 CNL-14-006 January 14, 2014 FPE 10 90 (extended to 120 days per electronic mail from NRC to TVA, dated February 6, 2014) CNL-14-025 March 14, 2014 E3-2 RAI Question Number Type of Response (days) Actual Date of Response FPE 11 90 (extended to 120 days per electronic mail from NRC to TVA, dated February 6, 2014) CNL-14-025 March 14, 2014 FPE 12 90 (extended to 120 days per electronic mail from NRC to TVA, dated February 6, 2014) CNL-14-025 March 14, 2014 FPE 13 120 CNL-14-025 March 14, 2014 Safe Shutdown Analysis (SSA) SSA 01 60 CNL-13-141 December 20, 2013 SSA 02 60 CNL-14-006 January 14, 2014 SSA 03 60 CNL-14-006 January 14, 2014 SSA 04 60 CNL-13-141 December 20, 2013 SSA 05 60 CNL-14-001 January 10, 2014 SSA 06 60 CNL-14-006 January 14, 2014 SSA 07 60 CNL-14-001 January 10, 2014 SSA 08 60 CNL-13-141 December 20, 2013 SSA 09 60 CNL-14-006 January 14, 2014 SSA 10 60 CNL-14-006 January 14, 2014 E3-3 RAI Question Number Type of Response (days) Actual Date of Response SSA 11 60 CNL-13-141 December 20, 2013 SSA 12 60 CNL-13-141 December 20, 2013 SSA 13 60 CNL-14-006 January 14, 2014 SSA 14 60 CNL-13-141 December 20, 2013 SSA 15 60 CNL-14-006 January 14, 2014 Programmatic (PROG) PROG 01 60 CNL-13-141 December 20, 2013 PROG 02 60 CNL-13-141 December 20, 2013 Fire Modeling (FM)

FM 01, part a 90 CNL-14-020 February 13, 2014 FM 01, part b.i 60 CNL-13-141 December 20, 2013 FM-01, part b.ii 60 CNL-13-141 December 20, 2013 FM-01, part b.iii 90 CNL-14-001 January 10, 2014 FM 01, part c 60 CNL-14-006 January 14, 2014 FM 01, part d.i 60 CNL-14-001 January 10, 2014 E3-4 RAI Question Number Type of Response (days) Actual Date of Response FM-01, part d.ii 90 CNL-14-001 January 10, 2014 FM 01, part e 60 CNL-13-141 December 20, 2013 FM 01, part f 60 CNL-14-001 January 10, 2014 FM 01, part g 90 CNL-14-020 February 13, 2014 FM 01, part h.i 60 CNL-14-001 January 10, 2014 FM 01, part h.ii 60 CNL-14-001 January 10, 2014 FM 01, part h.iii 90 CNL-14-020 February 13, 2014 FM 01, part i.i 120 CNL-14-020 February 13, 2014 FM 01, part i.ii 60 CNL-13-141 December 20, 2013 FM 01, part i.iii 90 CNL-14-001 January 10, 2014 FM 01, part i.iv 120 CNL-14-025 March 14, 2014 FM 01, part i.v 60 CNL-13-141 December 20, 2013 FM 01, part i.vi 60 CNL-13-141 December 20, 2013 FM 01, part i.vii 90 CNL-14-001 January 10, 2014 FM 01, part i.viii 60 CNL-13-141 December 20, 2013 E3-5 RAI Question Number Type of Response (days) Actual Date of Response FM 01, part j.i 60 CNL-13-141 December 20, 2013 FM 01, part j.ii 90 CNL-14-001 January 10, 2014 FM 02, part a 120 CNL-14-025 March 14, 2014 FM 02, part b 120 CNL-14-025 March 14, 2014 FM 02, part c 60 CNL-13-141 December 20, 2013 FM 02, part d 60 CNL-13-141 December 20, 2013 FM 02, part e 90 CNL-14-020 February 13, 2014 FM 03 60 CNL-14-001 January 10, 2014 Part b.i updated: CNL-14-025 March 14, 2014 FM 04 90 CNL-14-020 February 13, 2014 FM 05 60 CNL-14-001 January 10, 2014 FM 06 60 CNL-14-001 January 10, 2014 Probabilistic Risk Assessment (PRA)

PRA 01, part a 60 CNL-14-001 January 10, 2014 PRA 01, part b 60 CNL-14-006 January 14, 2014 E3-6 RAI Question Number Type of Response (days) Actual Date of Response PRA 01, part c 90 CNL-14-001 January 10, 2014 PRA 01, part d 90 CNL-14-020 February 13, 2014 PRA 01, part e 120 CNL-14-020 February 13, 2014 PRA 01, part f 120 CNL-14-025 March 14, 2014 PRA 01, part g 60 CNL-14-001 January 10, 2014 PRA 01, part h 90 CNL-14-020 February 13, 2014

Part ii updated: CNL-14-025 March 14, 2014 PRA 01, part i 60 CNL-13-141 December 20, 2013 PRA 01, part j 60 CNL-14-006 January 14, 2014 PRA 01, part k 60 CNL-14-001 January 10, 2014 PRA 01, part l 60 CNL-14-006 January 14, 2014 PRA 01, part m 60 CNL-14-001 January 10, 2014 PRA 01, part n 60 CNL-14-001 January 10, 2014 PRA 01, part o 120 CNL-14-025 March 14, 2014 E3-7 RAI Question Number Type of Response (days) Actual Date of Response PRA 01, part p 90 CNL-14-020 February 13, 2014 PRA 01, part q 60 CNL-13-141 December 20, 2013 PRA 01, part r 60 CNL-13-141 December 20, 2013 PRA 01, part s 90 CNL-14-020 February 13, 2014 PRA 01, part t 60 CNL-14-001 January 10, 2014 PRA 01, part u 60 CNL-13-141 December 20, 2013 PRA 01, part v 120 CNL-14-025 March 14, 2014 PRA 02 60 CNL-14-001 January 10, 2014 PRA 03 60 CNL-13-141 December 20, 2013 PRA 04 90 CNL-14-020 February 13, 2014 PRA 05 60 (extended to 120 days per electronic mail from NRC to TVA, dated January 9, 2014) CNL-14-025 March 14, 2014 PRA 06 60 CNL-14-001 January 10, 2014 PRA 07 60 CNL-14-006 January 14, 2014 PRA 08 60 CNL-14-001 January 10, 2014 E3-8 RAI Question Number Type of Response (days) Actual Date of Response PRA 09 60 CNL-14-001 January 10, 2014 PRA 10 90 (extended to 120 days per electronic mail from NRC to TVA, dated

February 6, 2014) CNL-14-025 March 14, 2014 PRA 11 60 CNL-14-006 January 14, 2014 PRA 12 120 CNL-14-025 March 14, 2014 PRA 13 60 CNL-14-006 January 14, 2014 PRA 14 60 CNL-14-001 January 10, 2014 PRA 15 90 (extended to 120 days per electronic mail from NRC to TVA, dated February 6, 2014) CNL-14-025 March 14, 2014 PRA 16 90 CNL-14-020 February 13, 2014 PRA 17 90 CNL-14-020 February 13, 2014 PRA 18 60 CNL-14-001 January 10, 2014 PRA 19 60 CNL-14-001 January 10, 2014 Part b updated:

CNL-14-020 February 13, 2014 PRA 20 120 CNL-14-025 March 14, 2014 E3-9 RAI Question Number Type of Response (days) Actual Date of Response PRA 21 60 CNL-14-001 January 10, 2014 PRA 22 60 CNL-13-141 December 20, 2013 PRA 23 60 (Part d extended to 120 days per electronic mail from NRC to TVA, dated January 9, 2014) Parts c, f, i, and l:

CNL-14-001 January 10, 2014 Parts a, b, e, g, h, j, and k: CNL-14-006 January 14, 2014 Part d: CNL-14-025 March 14, 2014 Radioactive Release (RR) RR 01 60 (extended to 120 days per electronic mail from NRC to TVA, dated

January 9, 2014) CNL-14-025 March 14, 2014 RR 02 60 (extended to 120 days per electronic mail from NRC to TVA, dated January 9, 2014) CNL-14-025 March 14, 2014

E4-1 ENCLOSURE 4 Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 Revisions to Proposed Modifications LAR Attachment S, Table S-2, provided the list of committed modifications associated with NFPA 805. As a result of continuing reviews of the modifications, TVA has revised two of the modifications and deleted four of the modifications as follows.

Item Proposed Modification Change Justification / Comment 26 Remove cable 0ABN6617 from tray DG. LAR, Attachment S, Table

S-2 is revised to delete

Item 26. The subject cable was

determined not to be located in Fire Area 05. Therefore, this modification is not required. 55 Install or modify mode switches at the RMOV BD for Shutdown Cooling Suction Valves (FCV-074-0002, 12, 25

and 36 for all 3 Units)

and Suppression Pool Suction Valves (3-FCV-074-0001, 12, 35) to

prevent spurious actuation during power operation for fires

external to the RMOV BD. LAR, Attachment S, Table S-2, Item 55, Proposed Modification is revised to replace "(3-FCV-

074-0001, 12, 35)" with "(3-FCV-074-0001, 13, 35)." The reference to valve FCV-074-12 in the proposed modification was a typographical error. 75 Electrically Isolate the Main Generator CT circuits from cables in

the Auxiliary Instrument Rooms. LAR, Attachment S, Table S-2, Item 75, Problem Statement is revised to delete "for the EHC system," and add "Fire damage to cables can result in loss of USST cooling."

Proposed Modification is

revised to "Re-route cables

for the Main Generator CT circuits away from the Auxiliary Instrument Rooms.

Install separate fuses in Main Relay Panel for USST coolers" Re-routing the cables provides better separation than providing isolation

devices.

Additional cables were added to the scope to further reduce the probability of losing offsite

power in the Auxiliary

Instrument Rooms (Fire

Areas 16-K, 16-M, 16-O).

E4-2 77 Install area wide Incipient Detection in Auxiliary Instrument Rooms. Delete the modification (see referenced RAI responses) The TVA responses to FPE RAI 10 (in Enclosure 1 to this letter) and FPE RAI 12, Part c (in Enclosure 1 to this letter) describe the deletion of Item 77 from LAR, Attachment S, Table S-2.

95 Re-configure the "43AR" switch for DG breakers 1818 (A),

1822 (B), 1812(C),

1838(D), 1842(3EB), 1832(3EC), and 1836(3ED) to isolate OTX, R1x, and ESTR signal cables from the trip circuit. LAR, Attachment S, Table

S-2 is revised to delete

Item 95. This same function of tripping

the Diesel Generator (DG) breaker can be accomplished

with an existing mechanical trip pushbutton also located on the breaker with the 43AR switch. Therefore, this modification is not required. 96 Replace Fire Pump pressure gauges 0-PI-26-46, -47, and -48 with gauges having a higher range of 0-300 psig. LAR, Attachment S, Table

S-2 is revised to delete Item 96. The existing gauges are 0-300 psig.

Therefore, this modification is not required.

The following table provides markups of the current LAR, Table S-2 wording to clarify the changes. Inserted text is underlined and deleted text is shown in strike-out.

Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed

Characterization 26 Med 1,2 Cable tray DG in Fire Area 05 contains abandoned, unqualified cable 0ABN6617 which adversely impacts fire modeling in that area. Remove cable 0ABN6617 from tray DG.

Yes Yes Risk is reduced by eliminating the unqualified cable in Fire Area 05. This unqualified cable could result in higher HRR's and the potential for self-ignited cable fires. 55 High 1,2,3 Fire damage to cables may cause spurious operation of RHR pump suction valves resulting in flow blockage, pump trips and valve interlocks.

Install or modify mode switches at the RMOV BD for Shutdown Cooling Suction Valves (FCV-

074-0002, 12, Yes Yes Risk is reduced by increasing the availability of suppression pool cooling.

E4-3 Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed

Characterization

Associated VFDRs: VFDR-16-0126 25 and 36 for

all 3 Units) and Suppression Pool Suction

Valves (3-

FCV-074-0001, 1213, 35) to prevent

spurious actuation

during power operation for fires external to the RMOV BD. 75 High 1,2,3 Fire damage to Main Generator CT circuits for the EHC system located in the Auxiliary Instrument Room can cause loss of

Main Transformers and loss of offsite power.

Fire damage to cables can result in loss of USST cooling.

Associated VFDRs:

Risk reduction modification not associated with

a specific VFDR. Electrically Isolate the Main Generator CT circuits from cables in the Auxiliary Instrument Rooms.Re-route cables for the Main Generator CT circuits away from the auxiliary instrument rooms.

Install separate fuses in Main Relay Panel for USST coolers Yes Yes Risk is reduced by preventing the LOOP due to a fire in the

Auxiliary

Instrument Rooms. 77 Med 1,2,3 Incipient Detection that covers general floor of the Auxiliary Instrument Install area wide Incipient Detection in Auxiliary Instrument Rooms. Yes Yes Risk is reduced by improving early indications of fire precursors for transient fire E4-4 Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed

Characterization Rooms is needed to improve fire detection and suppression probability for transient fire scenarios in the room.

Associated VFDRs: Risk reduction modification not associated with a specific VFDR. scenarios.

95 High 1,2,3 For fire scenarios in which the normal offsite power feeder breaker is manually aligned at the 4kV Shutdown Board to restore power, cable damage can clear control power fuses and prevent the DG breaker from opening.

Associated VFDRs: Risk reduction modification not associated with a specific VFDR. Re-configure the "43AR" switch for DG breakers 1818 (A), 1822 (B), 1812(C), 1838(D), 1842(3EB), 1832(3EC), and 1836(3ED) to isolate OTX, R1x, and ESTR signal cables from the trip circuit.

Yes Yes Risk is reduced by allowing isolation of OTX, R1x and ESTR signal cables from the trip circuit for the DG breakers.

96 Low Common Current Fire Pump pressure gauges do not have sufficient range to meet Replace Fire Pump pressure gauges 0-PI-26-46, -47, No Yes This is an NFPA 805 Chapter 3 compliance modification. This level of E4-5 Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed

Characterization NFPA 20 criteria.

Associated NFPA 805 Chapter 3 reference: 3.5.3 and -48 with gauges having a higher range of 0-300 psig. detail is not modeled in the PRA.

For consistency with the changes to LAR, Table S-2 discussed above, LAR Attachment A, "NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements," NFPA 805 Chapter 3 Reference 3.5.3 [Water Supply Pump Code Requirements] Compliance Basis is revised to delete the reference to Modification 96. Specifically, the last paragraph of the Compliance Basis section for NFPA 805 Chapter 3 Reference 3.5.3 is revised to state:

"Corrective actions were identified in the Code Compliance Evaluation. These corrective actions are identified in Modification 97 in Table S-2 of Attachment S and Implementation Item 17 in Table S-3 of Attachment S." After the Fire PRA is updated, the affected sections in the LAR will be revised to reflect the results of the updated Fire PRA quantifications. The revision to the LAR will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in response to remaining NRC Requests for Additional Information (RAIs).