ML13298A702

RAI Regarding LAR to Request Adopt National Fire Protection Association Standard 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (TAC Nos. MF1185 - MF1187)

ML13298A702
Person / Time
Site:	Browns Ferry
Issue date:	11/19/2013
From:	Farideh Saba Plant Licensing Branch II
To:	James Shea Tennessee Valley Authority
Saba D
References
TAC MF1185, TAC MF1186, TCA MF1187
Download: ML13298A702 (48)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 19, 2013 Mr. Joseph W. Shea Corporate Manager- Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 11 01 Market Street LP 3D-C Chattanooga, TN 37402-2801

SUBJECT:

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)

Dear Mr. Shea:

By letter dated March 27,2013, as supplemented by letter dated May 16, 2013, Tennessee Valley Authority (the licensee, TVA) submitted a license amendment request to transition the fire protection licensing basis at the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (BFN), from Title 10 of the Code of Federal Regulations (CFR), Section 50.48(b), to 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants."

The U.S. Nuclear Regulatory Commission (NRC)'s staff from the Fire Protection Branch, the Probability Risk Assessment Licensing Branch, and the Health Physics & Human Performance Branch reviewed the information provided by TVA, participated in an audit at the BFN site from September 9 - 13, 2013, and have determined that additional information is needed to complete the review. On September 26, 2013, the NRC staff forwarded, via an electronic mail, a draft of the request for additional information (RAI) to the TVA staff. On October 23, 2013, the NRC staff and representatives of TVA held a conference call to provide the licensee with an opportunity to clarify any portion of the draft RAI and discuss the timeframe for which TVA may provide the requested information. The finalized RAis are found in Enclosure 1 of this letter. As a result of that conference call, the NRC staff and Jerry Jones of your staff agreed to the schedule in Enclosure 2 of this letter for providing responses to the finalized RAis.

J.Shea Please contact me at 301-415-1447, or via email at farideh.saba@ nrc.gov, if you have any questions.

Sincerely, j-ot Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296

Enclosures:

1. Request for Additional Information

2. Licensee's Response Schedule cc w/enclosures: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT-WATER REACTOR GENERATING PLANTS TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 DOCKET NOS. 50-259, 50-260, 50-296 Fire Protection Engineering (FPE)

FPE RAI 01 License Amendment Request (LAR) (Agencywide Documents Access and Management System (ADAMS) Accession Number ML13092A392), Attachment A, Table B-1, Section 3.3.6, Roofs, indicates that testing for a Class I Factory Mutual rating is equivalent to testing for a Class A National Fire Protection Association (NFPA) 256, "Standard Methods of Fire Tests of Roof Coverings," rating. Provide technical justification for this equivalency statement.

FPE RAI 02 LAR Attachment A, Table B-1, Section 3.4.1 (c) indicated "complies" using the exception for the incident commander and at least two brigade members having sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety systems. Provide the details of the minimum "training and knowledge" the incident commander receives. For example, identify whether the incident commander will be a licensed or non-licensed operator. Describe how the incident commander is expected to function with the fire brigade.

FPE RAI 03 LAR Attachment A, Table B-1, Section 3.5.5, Water Supply Pump Separation, states that "Complies by Previous Approval" is the compliance strategy. The documentation for such approval seems to be a general statement indicating that the safety evaluation approved the submitted program. Provide more details, specifically regarding the description of fire pump separation from each other, as well as the rest of the plant, and the specific Nuclear Regulatory Commission (NRC) approval chain of correspondence and approvals that lead to this conclusion.

Enclosure 1

FPE RAI 04 LAR AttachmentS, Table S-3, Implementation Item 44, states "Implement corrective actions as required to ensure that pressure is maintained in the fire protection system during normal operation without using a fire pump." Provide a more detailed description of this issue, and the intended/proposed resolution. If necessary, explain the type of modification to the system this would entail. If a modification is indicated, update LAR AttachmentS, Table S-2 accordingly.

FPE RAI 05 LAR Attachment V, Fire PRA [Probabilistic Risk Assessment] Quality,Section V.2.5, states that

" ... in some cases, the 1-hour rated ERFBS [electrical raceway fire barrier system] 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." Describe the method used from a regulatory standpoint for resolution of these identified ERFBS conditions including the use of variance from deterministic requirements (VFDRs) for identification, resolution, and evaluation of the safety margin and defense-in-depth (DID). Describe how these ERFBS resolutions were identified (e.g., Table C-2). If VFDRs were not identified, provide justification for not using the fire risk-evaluation (FRE) process.

FPE RAI 06 The NRC staff noted that 1-hour rated fire wrap is used for exclusion of intervening combustibles within 20 foot separation zones. Describe how this fire protection feature was transitioned to the new risk-informed, performance-based (RI/PB) fire protection program (FPP).

FPE RAI 07 The NRC staff noted that the plant FTP documents describe all 1-hour rated fire wrap as Thermo-Lag. Identify all the different ERFBSs (e.g., 3-M, Hemyc, etc.) in the NFPA Standard 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants" (NFPA 805), 2001 Edition, post-transition configurations.

FPE RAI 08 The residual heat removal (RHR) pump/heat exchanger rooms lack automatic suppression systems and it was noted that water curtain and draft stop were previously installed in these rooms to separate them from the rest of the fire area. The NRC granted an Appendix R exemption for the lack of suppression and in the LAR, this exemption is requested to be carried forward. Although this "water curtain" is addressed in LAR, Attachment C, Table C-2, the "draft stop" is not listed. It appears that the "draft stop" was part of the original exemption and should be carried over as a fire protection feature. Provide justification for this apparent discrepancy.

Identify any other locations where "draft stops" are credited.

FPE RAI 09 The NRC staff noted that special administrative controls will be in place to limit transient fires to 69 kilowatts (kW) in Fire Compartments 5, 9, 16A, 16K, 16M, and 160. The fire hazards

analysis describes combustibles in Fire Area 16 to be cable insulation, foam plastics, oil, lube oil, wood, paper, cloth, polyester and neoprene, charcoal, rubber, cellulose, styrofoam, and plywood. The assumption in LAR Attachment V,Section V.2.3, for 69 kW fires states that all fire compartments that were credited for a reduced heat release rate (HRR) will be subject to strict combustible controls (fire compartments designated as "No Combustible Storage") and paper, cardboard, scrap wood, rags and other trash shall not be allowed to accumulate in the area.

Provide a more detailed description of the control of combustibles. For Fire Compartment 16A (Main Control Room (MCR) for Unit 1 and Unit 2), provide details with regard to restricting the accumulation of paper, books, binders, trash, etc. Describe any special administrative controls that are expected to control transient fires to 69 kW, as opposed to the HRR (i.e., 317 kW) in Table G-1 of NUREG/CR-6850 and Electric Power Research lnsitute (EPRI) 1011989, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities."

FPE RAI10 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 (ADAMS 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.

FPE RAI11 Incipient detection is described in the LARin 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 AttachmentS, 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 AttachmentS, 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, 16-0, 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.

FPE RAI12 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.

FPE RAI13 In LAR Attachment I, Table 1-1 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, standby gas treatment building, cooling towers are included in LAR AttachmentW, Tables W-8, W-9, and W-10, but listed as part of the YARD in Table 1-1. 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 (f1) CDF by unit) between Tables W-8, W-9, and W-10.

c) In LAR Attachment W, Tables W-8, W-9, and W-1 0, 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.

Safe Shutdown Analysis (SSA)

SSA RAI 01 LAR Section 4.2.1.1 identifies that a "representative population" of motor operated valves (MOVs) was evaluated to address potential pressure boundary concerns in a technical report entitled, "NFPA 805 MOV Stall Thrust Pressure Boundary Evaluation," reviewed during the on-site audit. Explain how the failure mode was evaluated in the FPRA. Provide a list of those MOVs that may be susceptible to Information Notice (IN) 92-18, "Potential for Loss of Remote Shutdown Capability During a Control Room Fire," failure, and also may be required to be operated as a recovery action (RA). Describe the methods by which operator feasibility was resolved for those susceptible to IN 92-18 failure mode.

SSA RAI 02 LAR Section 4.2.1.1 currently states that an open secondary circuit on a current transformer creates the need to perform an FRE for a secondary fire. The NRC staff noted that TVA calculations reviewed during the audit, indicate that open circuit secondary fires are possible.

VFDRs for open secondary circuits on current transformers are identified in LAR Attachment C, Table C-1. For a fire in the control room (Fire Area 16), VFDR 16-0243 indicates a potential open circuit secondary fire at "locations outside the control room," but resolves the VFDR by risk evaluation and no further action required. Provide more details regarding the following:

a) Explain the methodology used for CR abandonment, loss of control with alternate shutdown, or if operators remain in the CR to determine the risk contribution to the CR fire, and the justification for no further action required. Describe the safety margin and DID elements that Table B-3 indicates were satisfied.

b) Describe how the secondary fire was considered in this analysis.

c) Provide a list of locations for these secondary fires.

d) Provide a list of RAs for a fire in the CR required at any of the locations or panels susceptible to secondary fires.

e) Describe what components, if any, may be involved in the secondary fire, and are also required for an RA. Describe how this was accounted for in the risk analysis and the feasibility for RAs.

SSA RAI 03 LAR Attachment C, Table C-1, identifies approximately 18 fire areas where VFDRs identify the potential of ignition of "secondary fires at locations outside" the fire area being analyzed. These are represented as "common enclosure" circuit issues. Provide more details regarding the following:

a) Explain the methodology used to determine the risk contribution to fire, and the justification for those with no further action required. Describe the safety margin and DID that Table 8-3 indicates were satisfied.

b) Describe how the secondary fire was considered in the analysis.

c) Provide a list of locations for these secondary fires for each VFDR.

d) Provide a list of RAs for a fire in the CR required at any of the locations susceptible to secondary fires.

e) Describe what components, if any, may be involved in the secondary fire, and are also required for an RA. Describe how this was accounted for in the risk analysis and the feasibility for RAs.

SSA RAJ 04 LAR Attachment 8, Table 8-2, Section 3.5.2.3 "Circuit Failures Due to Hot Short," identifies that treatment of hot shorts on ungrounded circuits "aligns with intent" of Nuclear Energy Institute (NEI) 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 1. The LAR states that "two types of cable hot short conditions are considered to be of sufficiently low likelihood that they are not assumed credible, except for analysis involving high/low pressure interface components in accordance with NEI 00-01." These hot short exceptions are 3 phase alternating current power circuit cable-to-cable proper phase sequence faults and 2-wire ungrounded direct current (DC) circuit cable-to-cable proper polarity faults. Provide a more detailed justification for this deviation from the guidance of NEI 00-01, Revision 1.

SSA RAJ 05 LAR Section 4.2.1.2 states that the at-power analysis is based upon the ability to achieve and maintain hot shutdown conditions (Mode 3).

a) Describe the specific capabilities that will be required to meet the performance criteria beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. One example is that "Liquid nitrogen tanks for safety relief valve pneumatics are sufficient for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after which sources may be required."

b) Describe any system or component capacity limitations and time-limited actions needed to replenish systems, make repairs, or otherwise maintain safe and stable conditions, (e.g., DC battery power, diesel generator fuel, water inventory, etc.).

c) Describe whether there are any actions to recover/repair nuclear safety capability assessment (NSCA) equipment to sustain safe and stable conditions. Describe the resource (staffing) requirements and timing of these actions.

d) Describe how the feasibility of the actions in Items band c (above) was evaluated or addressed.

e) Provide a more detailed discussion (qualitative or quantitative) of the risk of failure of the actions necessary to sustain safe and stable conditions beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

f) LAR Section 4.2.1.2 states that off-site power is available to provide long-term power.

Explain the basis for this statement. Describe what repairs are necessary to establish this availability.

SSA RAI 06 LAR AttachmentS, Table S-2, Items 35 and 35a identify the proposed installation of a new emergency high pressure makeup pump for each unit. Provide a more detailed description of the proposed modification. Include in the description, equipment and systems that will be necessary to support the operation of this pump (electrical power, water supply, control valves, fuel oil supply, etc.). Describe the proposed operation of the pump including local and remote operations, and how it is intended to function during safe shutdown and/or alternate shutdown.

Describe the performance goals it would support. With regard to the current transformer analysis for secondary fires, include in this description the potential impact of secondary fires for the new pump power supplies.

SSA RAI 07 LAR AttachmentS, Table S-3, Items 27, 28, 29, 30, 31, 32, and 33, indicate the need to develop or revise post-fire response procedures, identify required tools during procedure validation and verification, document new staffing requirements, conduct training, and update the human reliability analysis (HRA) and FPRA model. Additionally, the methods of alternate and safe shutdown process will be transitioning from the "self-induced station blackout" (SISBO) method to the NFPA 805 post transitioned fire safe shutdown procedures. There are also new success paths including an emergency high pressure makeup pump, and the condensate and booster pump feedwater injection success paths. RAs have not been fully evaluated for feasibility because procedures have not been completed. These outstanding work activities present a potential for significant impacts to the current NSCA presented in the LAR. Provide justification regarding the confidence that the results presented in the LAR will not significantly change.

SSA RAI 08 LAR Attachment D describes the methods and results for non-power operations (NPO) transition. Provide the following additional information:

a) At a high level, identify and describe the changes to outage management procedures, risk management tools, and any other documents resulting from incorporation of key safety functions identified as part of NFPA 805 transition for NPO for high risk periods. Include changes to any administrative procedures such as control of combustibles, hot work, fire system operability, staffing, and ignition sources.

b) Provide justification for not performing the NPO analysis for a fire in the MCR. In lieu of this analysis describe the provisions made in the event of a fire in the MCR during high risk evolutions. Include in this response all rooms within Fire Area 16, not just the MCR proper.

SSA RAI 09 For breaker fuse coordination, describe whether cable length was credited as additional impedance in the study necessary to meet maximum available fault current to demonstrate coordination with upstream power supply breakers. If this was used, describe how this length was factored into the potential impact to the FPRA. For establishing targets in the zone of influence (ZOI) describe how cable lengths were considered and provide any justifications required for the FPRA.

SSA RAI10 NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 1, and NFPA 805 identify circuit failure types (hot short, open circuit, short-to-ground) that need to be considered in the circuit analysis. LAR Attachment B, Table B-2 indicates alignment with the guidance of NEI 00-01, Revision 1. The NRC staff noted that certain circuit configurations credit special "shorting switches" as protection against hot shorts. For those components required to meet the nuclear safety performance goals of NFPA 805, provide a list of components that will rely on this protection scheme. Include the equipment type (MOV, SOV (solenoid operated valve), air operated valve, etc.), type of control circuit (grounded or ungrounded, control power transformer (CPT) or non-CPT), fire locations where credit was taken, and any justification for relying on the circuit not failing open in a fire. Identify where any of these circuits are being modified using this configuration and indicate which of these are listed in LAR Attachment S.

SSA RAI11 LAR Attachment B, Table B-2, Section 3.2.1.2, "Fire Damage to Mechanical Components,"

indicates that fire-induced damage to mechanical components subjected to a fire is not considered credible. However, NEI 00-01, Revision 1 states that instrument tubing with brazed, soldered joints, or other heat sensitive materials should not be included in this non-failure assumption. Describe how the failure of brazed or soldered joints in a fire was considered in the NSCA.

SSA RAI12 LAR Attachment S, Table S-2, Items 9a through 9f, describe 10 RHR valves being modified to prevent water hammer. Provide more detail of the modifications with regard to specific circuit changes and how these modifications will prevent spuriously opening the valves. Include in the description the use of insulated shield cables and how they function to protect the valve circuit.

SSA RAI13 The LAR states that DID and safety margin guidance from NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under [Title 10 of Code of Federal Regulations] 10 CFR 50.48(c)," and Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,"

are used in the risk evaluations when resolving VFDRs. LAR Attachment C, Table C-1 identifies FPP systems and features by fire area credited for DID. DID and safety margin are stated as being evaluated as part of the area-by-area FRE. However, the specific criteria used to

determine adequate DID and safety margin is not provided in the LAR. Provide additional discussion of the methods and criteria for evaluating DID and safety margins and summarize the results as required by NFPA 805, Section 4.2.4.2.

SSA RAI14 Cable routing information being used for the FPRA is described as primarily based on the previous Appendix R safe shutdown data. Certain measures were developed that validate the accuracy of this routing information. Provide a detailed description of these processes and justification of how they were used to validate the accuracy of this data. Include in that description the efforts and scope to 'nodalize' raceways for more accurate location information.

Describe how walkdown information was used to assist in this validation.

SSA RAI15 The shutdown method is transitioning from SISBO to fire protection procedures using the performance-based analysis of NFPA 805. Provide a description of the steps being undertaken to ensure a complete and smooth transition including RAs (including feasibility), modifications, training, and operating procedures.

Programmatic (PROG)

PROG RAI 01 Based on the NRC staff's review of the LAR and associated documentation, it was determined that the LAR did not provide the information needed for the NRC staff to evaluate what changes will be made to the FPP related training to incorporate NFPA 805 requirements.

Describe the changes that are planned to the training program to prepare the plant staff for implementing the new risk-informed, performance-based FPP as part of the NFPA 805 transition process.

PROG RAI 02 NFPA 805, Section 2.7.3.4, "Qualification of Users", states that cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

Describe how the training program will be revised to support the NFPA 805 change evaluation process, including positions that will be trained and how the training will be implemented (e.g.,

classroom, computer-based, reading program).

Fire Modeling (FM)

FM RAI 01 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:

a) The NRC staff identified the possibility that non-cable intervening combustibles were missed in fire areas of the plant. Provide information on how non-cable intervening combustibles were identified and accounted for in the fire modeling analyses.

b) Regarding fire propagation in cable trays:

i. Describe how the ignition of and subsequent flame spread and fire propagation in (stacks of) horizontal cable trays and the corresponding HRR of cables were calculated.

ii. Explain how cables with Flamemastic 77 coating and cable tray with covers were treated in the fire modeling calculations.

iii. Explain if and how the presence of holes in cable tray covers was accounted for.

c) During the audit the NRC staff learned that wall and corner effects were not accounted for in the HGL and multi-compartment fire modeling analysis. Provide technical justification for not considering wall and corner effects in the fire modeling calculations.

d) The HRR of electrical cabinets throughout the plant is based on the assumption that they are closed and contain multiple bundles of unqualified cable (Case 4 in Table E-1 of NUREG/CR-6850, Vol. 2).

i. During the audit the NRC staff noted that the door of an electric cabinet in the area behind the Unit 1 MCR horseshoe was partially open. Provide justification for the assumption that there are no open cabinets in the plant.

ii. During the audit the NRC staff noted that several electrical cabinets in the auxiliary instrument rooms have Plexiglass doors. Describe and technically justify the HRR that was used for these cabinets.

e) Explain if the guidance in Appendix Min NUREG 6850, Vol. 2 was used to determine the damage due to high energy arcing faults, or provide technical justification if a different approach was used. In addition, provide technical justification for the assumption that a fire following a high energy arcing fault event has an HRR of 211 kW for duration of 20 minutes, followed by a 20-minute decay.

f) During the audit the NRC staff noted transient combustibles in several areas that may not have been considered in the fire modeling analyses (e.g., upholstered chairs, carts with combustibles and computer equipment in several areas; large quantities of resin near the Unit 1 oil tanks in the turbine building; self-contained breathing apparatus cases, paper in cardboard boxes, a copier and printers in the control room areas, etc.). Describe whether the fire modeling analyses that were conducted are bounding for scenarios that involve these transient combustibles. In addition, explain how it is ensured that the model assumptions in terms of transient combustibles in a fire area or zone will not be violated during and post-transition.

g) During the audit the NRC staff observed some mechanical equipment in the Unit 3 CSR.

Describe whether fire scenarios involving oil from this equipment were considered in the fire modeling analyses for this fire area.

h) Specifically regarding the use of the algebraic models:

i. Explain how the elevation of ignition source fires was determined, and describe whether the assumptions are consistent with plant conditions or if they lead to conservative ZOI and HGL temperature estimates.

ii. Provide technical justification for using a fixed Froude number value of one to determine the diameter of the fire as opposed to the actual diameter of the fire. Determine the range of fire diameters that correspond to a Froude number of one for the HRRs of the fires that were considered in the fire modeling analyses and show that this range is reasonably consistent with the dimensions of transient combustibles in the plant.

iii. During the audit the NRC staff observed that the assumption in the HGL calculations of a 10 x 10 tt or 3 x 7 ft natural ventilation opening may not be consistent with plant conditions. Provide technical justification for the exclusive use of the McCaffrey-Quintiere-Harkleroad method for the HGL calculations (as opposed to Seyler's method for closed compartments) and for the assumed vent dimensions.

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

i. Provide technical justification for the assumption that the polycarbonate ceiling panels in the MCR do not significantly contribute to a fire. Since the ceiling covers a very large area, there is a concern for rapid flame spread across the surface of the panels and burning droplets or lumps of melted polymer.

ii. Provide technical justification for using a soot yield of 0.12 for the cables in the MCR FDS simulations. In addition, explain what value (or values) was (were) used for the heat of combustion of cables in the MCR. Describe whether the soot yield and heat of combustion values that were used in the analysis result in conservative estimates of the soot generation rate.

iii. Confirm that the 10 in. by 6 in. holes that were created in the polycarbonate ceiling are consistent with the actual openings to the interstitial space, or explain why the assumed hole dimensions are conservative in terms of CR abandonment.

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./5x) 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./5x values outside the NUREG-1824 recommended range.

v. Provide technical justification for assuming a transient fire elevation of 1 meter above the floor, and for choosing an area so that the Froude number is within the NUREG-1824 validation range (as opposed to the actual transient size).

vi. FDS "devices" (temperature, heat flux, and optical density) were placed at different locations around the MCRs. Describe the basis for choosing these locations.

vii. Provide technical justification for assuming that transient fires in the MCR reach peak HRR in 8 minutes.

viii. FDS simulations were performed with cabinet and transient fires located at four different locations. Describe the technical basis that was used for choosing these locations.

j) Specifically regarding the MCA:

i. Describe the criteria that were used to screen multi-compartment scenarios based on the size of the exposing and exposed compartments.

ii. In the MCA it is assumed that there is a natural ventilation opening of 10 x 10ft or 3 x 7 ft in the exposing compartment, and that there are no natural ventilation openings in the exposed compartment. During the audit the NRC staff observed that these assumptions may not be consistent with plant conditions. Provide technical justification for the vent dimensions in the exposing and exposed compartments assumed in the CFAST multi-compartment calculations.

FM RAI 02 American Society of Mechanical Engineers/American Nuclear Society (ASMEIANS) Standard RA-Sa-2009, "Addenda to ASMEIANS 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 compoenents. 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.

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 qualified 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.

c) Describe how cable tray covers and conduits affect the damage thresholds that were used in the fire modeling analyses. In addition, explain how holes in cable tray covers were treated in this respect.

d) Explain what damage thresholds and associated time delays were used in the fire modeling analyses for cables coated with Flamemastic 77.

e) During the audit the NRC staff noted that several electrical cabinets in the auxiliary instrument rooms have Plexiglass doors with gaskets. The Plexiglass doors of two cabinets were partially open. The cabinets with Plexiglass doors appear to contain sensitive electronic equipment. Describe the damage criteria that were used for this equipment.

Provide technical justification if the thresholds that were used in the fire modeling analyses are different from those recommended in Section H.2 of NUREG/CR-6850, Vol. 2.

FM RAI 03 NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). Reference is made to LAR Attachment J, "Fire

Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used. Furthermore, LAR Section 4.7.3 "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" states that "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

Regarding the V& V of fire models:

a) Explain how the workbooks that were developed to calculate the ZOI and HGL temperatures were verified.

b) Seyler's method to estimate the HGL temperature in closed compartments was used in the MCA, but there is no discussion of this model in Attachment J of the LAR and the V& V report (reference 16 in the LAR).

i. Describe how Seyler's method (as implemented) was verified.

ii. Provide the V& V basis for Seyler's method.

iii. Provide technical details to demonstrate whether the method has been applied within the validation range of input parameters, or to justify the application of the equations outside the validation range reported in the V&V basis documents.

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

FM RAI 04 NFPA 805, Section 2.7.3.3, "Limitations of Use," states: "Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verifications and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method."

LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805,"

states that "Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were and are used with the same limitations and assumptions supported by the V&V for the methods as required by Section 2.7.3.3 of NFPA 805."

Regarding the limitations of use:

a) Algebraic models cannot be used outside the range of conditions covered by the experiments on which the model is based. NUREG-1805, "Fire Dynamics Tools (FDTs),"

has a section on assumptions and limitations that provides guidance to the user in terms of proper and improper use for each FDT. There is general discussion of the limitations of use for the algebraic equations that has been utilized for hand calculations. It is not clear, however, how these limitations were enforced on the individual fire areas or for the multi-compartment analysis. Provide a description of how the limit of applicability was determined for each fire area.

b) Identify uses, if any, of CFAST outside the limits of applicability of the model and for those cases explain how the use of CFAST was justified. Include a list of areas, zones and scenarios for which CFAST was used to confirm HGL development.

c) Identify uses, if any, of FDS outside the limits of applicability of the model and for those cases explain how the use of FDS was justified.

FM RAI 05 LAR Section 4.5.1.2, "Fire PRA states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together. Furthermore, LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states that "For personnel performing fire modeling or FPRA development and evaluation, TVA will develop and maintain qualification requirements for individuals assigned various tasks. Position specific guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work."

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

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

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

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

FM RAI 06 LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" states that "Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and Fire PRA development used to support performance-based approach."

Regarding the uncertainty analysis for fire modeling:

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

b) Describe how the "model" and "completeness" uncertainties were addressed for this application and accounted for in the analyses. NUREG-1934, "Nuclear Power Plant Fire Modeling Analysis Guidelines," provides guidance on quantifying model/completeness uncertainty.

c) An example of these uncertainties is the effect of ignoring the contents of a compartment on the HGL calculations. Cabinets and cable trays reduce the effective volume of a compartment, but also act as a heat sink. Hence, how ignoring these contents affects the HGL temperature is not clear. Explain how the corresponding uncertainties were accounted for, or show that ignoring compartment contents leads to conservative HGL temperatures.

Probabilistic Risk Assessment CPRA)

PRA RAI 01 - Fire PRA Facts and Observations (F&Os)

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:

a) F&O 2-47 against CF-A 1:

This F&O 2-47 disposition explains that additional components were added to what was originally a small population of components analyzed for their circuit failure mode likelihood.

A larger scope (more than 200 components) was analyzed to address this F&O. SR CF-A 1 requires that conditional failure probabilities be assigned to risk significant contributors. The resolution to this F&O stated that the top 50 scenarios were reviewed to identify potentially important components. State the percentage of risk that these scenarios represent, and summarize how this review provides confidence that components associated with risk-significant spurious operations have been identified.

b) F&O 4-2 against CS-81:

The disposition to F&O 4-2 states that it has been addressed by including in the FPRA the risk contribution of secondary fires due to the presence of a specific type of current transformer in electrical buses. In the NSCA and FPRA, current transformer cables are included in the analysis. The corresponding cables were routed so that ignition sources that may affect the cables are identified, an ignition frequency is assigned, and adjacent targets can be identified for secondary fires as they are postulated. However, there is insufficient information to understand how that analysis was performed. Provide a more detailed description of the method used to quantify the risk associated with this event including assumptions made about extent of the impact from the secondary fire. Because these secondary fires occur in fire areas beyond the originating fire, include in the explanation the impact to other components and to recovery actions required to be conducted in the areas of secondary fires. Also, explain whether the feasibility of those actions was considered.

c) F&O 1-7 against FSS-A4, FSS-A5 and FSS-A6:

Appendix D of the MCR Analysis categorizes a number of detached panels, which appear to be associated with offsite power (i.e., 9-23-1 through 9-23-8), as part of the main control board rather than as electrical panels. In addition, the disposition to the F&O notes that electrical panels 3-PNLA-009-0023CD and 3PNLA-009-0023BA were reclassified as part of the main control board following the peer review. Provide the basis for including these electrical panels as part of the main control board using guidance provided in Supplement 1 to NUREG/CR-6850.

d) F&O 2-38 against HRA-A2:

The disposition to this F&O states that the final fire procedures are not yet available to complete and verify the FPRA HRA. Owing to the number of modeling assumptions noted in the Post-Fire HRA report, it is unclear whether the HRA performed for the FPRA is representative of the post-transition, non-SISBO fire response procedures and unclear to what extent these procedures have been developed. Additionally, numerous other F&Os (e.g., 2-39, 2-41, 2-50, 4-3, 4-12, and 4-21) are identified as open items or remain only partially addressed due to the incompleteness of fire procedures. As a result, describe the extent to which the post-transition, non-81880 fire response procedures have been developed as well as the degree to which they have been used to support the FPRA HRA.

e) F&O 2-39 against HRA-83:

The F&O identifies the lack of human failure event (HFE) definition as a "systematic issue."

The disposition indicates that the fire HRA has undergone significant enhancements since the peer review. As a result, address the following:

i. Describe how the detailed analysis performed for the FPRA adequately addresses performance shaping factors (including cues, applicable procedures, complexity of response, training, accessibility, environment, availability of resources, special requirements, and stress), and clarify the basis used to inform these factors.

ii. The HRA calculator files provided in Attachment D of the Post-Fire HRA report indicates that "dummy or "pseudo" values of T delay for operator actions were established to determine proper sequencing; however, this does not appear to explain the timing developed for actions to calculate their individual human error probabilities (HEPs).

Explain the basis for timing estimates (e.g., T8 w, Tdelay, T1 12, T M, etc.) utilized in the detailed analysis of individual actions.

iii. Due to the cited lack of HFE definition, clarify how factors that influence the level of dependency between HFEs (including intervening success, manpower, shift changes, common cognition, timing between cues, time required to complete actions, stress, and location) were addressed by the dependency analysis, and clarify the basis used to inform these factors.

iv. Explain how the above performance shaping and dependency factors are employed to address the impact of the fire on operator response.

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.

g) F&O 2-55 against FSS-E3:

According to the disposition provided for F&O 2-55, a quantitative characterization of the parameters used in the fire modeling of significant fire scenarios has not been completed.

Identify and characterize sources of uncertainty and related assumptions associated with parameters used for the modeling significant fire scenarios (e.g., a mean value and statistical representation uncertainty intervals). Provide a discussion of the criteria utilized to assess their importance to fire risk, and for those determined to be key sources of uncertainty for the FPRA, provide a sensitivity study of their impact on fire risk results (i.e.,

CDF, LERF, ~CDF and ~LERF). If no more than a qualitative characterization is performed, provide detailed justification describing why meeting FSS-E3 below CC-II is acceptable for this application.

h) F&O 2-56 against FSS-F3:

The disposition to this F&O seems incomplete. Address the following:

i. FSS-F3 requires the completion of a quantitative assessment of unscreened scenarios.

Describe the basis for only performing a qualitative assessment of the CSR scenario.

ii. The catastrophic turbine/generator (T/G) fire postulated for PAU 26A does not appear to be consistent with Table 0-2 of NUREG/CR-6850. In particular, a 0.025 conditional probability is utilized; however, the frequency of the scenario is not representative of the

sum of Bins 33, 34, and 35, but that of Bin 35 alone. Revise T/G scenarios to be consistent with NUREG/CR-6850 guidance, or alternatively provide a sensitivity analysis of the impact of this inconsistency on the risk results (i.e., CDF, LERF, ~CDF and

~LERF).

iii. Discuss whether a severe T/G oil fire was considered for its impact on structural elements in addition to a catastrophic T/G fire.

i) F&O 2-57 against FSS-D7:

A disposition to this F&O does not appear to be provided. Describe the credit given in the FPRA to incipient fire detectors in panels as well as the resulting impact on fire scenario development (e.g., damage end states), and clarify the resolution to the peer review observation that incipient detection credit should be removed from transient fire scenarios.

Include explanation of whether the incipient detection is used only to limit damage to targets outside the cabinet and not used to limit damage inside the cabinet where the detection would be installed. If incipient detection was used to limit damage inside the cabinets where detection would be installed, as opposed to limit damage to targets outside the cabinet, provide the impact on CDF, LERF, ~CDF, and ~LERF of removing the credit for this deviation from acceptable methods.

j) F&O 3-13 against FSS-D3:

The disposition to this F&O states that credit for automatic and manual suppression is applied to each damage state based on the suppression determined to have been activated.

Clarify how the time to detection and suppression system activation was determined for both seeping and detailed fire scenarios. Include in this clarification an assessment of fire detection and suppression systems effectiveness in the context of fire scenarios analyzed.

Additionally, discuss the process by which target sets are selected in developing the damage states of a detailed fire modeling scenario. Include in this discussion how target set damage times, hot gas layer effects, and secondary combustibles influence this selection process.

k) F&O 4-3 against SC-B1:

The disposition to this F&O states that "MSO scenario 1a is closed as the decision has been made on how it will be addressed." Multiple spurious operation (MSO) 1a could prevent a full scram. The NFPA 805 Multiple Spurious Operation Review report indicates that emergency operating instructions will be modified to instruct operators to take actions, both internal and external to the MCR, in case of the inability to scram; however, as stated in Appendix 3, "this action was not modeled in the FPRA since ATWS [anticipated transients without scram] is not modeled." Since fire induced failure could prevent a full scram it is not clear why this failure is not modeled in the FPRA. Provide further justification for not modeling this MSO scenario and related recovery actions in the FPRA.

I) F&O 4-17 against PRM-85:

The disposition to this F&O does not address the peer review team's observation regarding treatment of detrimental actions and identification of new fire-induced accident sequences.

Provide the following:

a. Discussion of the extent to which success of fire-specific actions were considered as a potential failure mode of components modeled in the FPRA model, including any new resulting accident sequences as appropriate.

b. Description of the review performed to identify new accident sequences, including timing considerations not in the internal events. Include in this description a disposition of the examples identified by the F&O.

m) F&O 4-18 against AS-83:

The disposition to this F&O only addressed flooding resulting from fire-induced interfacing system loss-of-coolant accidents (LOCAs). Summarize the consideration given to other potential fire-induced flooding scenarios (e.g., location, extent and consequences), and describe how such scenarios were treated in the FPRA.

n) F&O 4-28 against FQ-D1:

Describe the FPRA model LEAF-reducing refinements discussed in the disposition to this F&O.

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

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

p) F&O 5-6 against IGN-A4:

The disposition notes that a review of plant-specific fire experience was performed using guidance in NUREG/CR-6850 and that only one potentially challenging fire event was identified. However, based on the at-power fire events listed in Attachment 8 of the Fire Ignition Frequency report, the basis for classifying events as non-challenging remains unclear from the descriptions provided. During the audit, the NRC staff reviewed a small sample of past events and determined that at least one of these (i.e., Problem Evaluation Report (PER) 141331) was not classified as a potentially challenging event by TVA but should have been according to the criteria in NUREG/CR-6850, Appendix C (i.e., the event identified by PER 141331 involved active intervention by non-hot-work plant personnel to suppress the fire). Lastly, a review of the "Unit/Status" column appears to only indicate two

modes (i.e., power operation and shutdown); however, NUREG/CR-6850 indicates that events occurring during low-power operations, such as plant start-up, should be considered as well. Identify whether any of the fire events not classified as potentially challenging by TVA were classified in the updated EPRI fire event database (i.e., EPRI 1025284) as potentially challenging. If additional fire events were determined to be potentially challenging, discuss each of these events. Provide an assessment of the impact on the risk results provided in the LAR, Attachment W, or provide justification for why a Bayesian update is not required.

q) F&O 5-10 against IGN-A7:

In regard to the specific example cited by the F&O, the ignition source's contribution to Bin 15 appears to have been adjusted from 2 to 15; however, based on a review of Attachment 4 to the Fire Ignition Frequency report, its contribution to Bin 16a appears to have remained at 2. Clarify the discrepancy between these two counts.

r) F&O 5-11 against IGN-A7:

The disposition to this F&O notes that junction boxes were excluded as a fire ignition source. Describe the treatment of junction boxes in the FPRA as ignition sources, and clarify whether this treatment is consistent with FAQ 13-0006.

s) F&O 5-18 against FSS-C4 and F&O 8-3 against FSS-C1:

The disposition to F&O 5-18 appears to indicate that the risk contribution from the fraction of fires for an ignition source that do not propagate to targets beyond the ignition source was only quantified if the complement fraction of fires that do propagate result in a CDF contribution above 1.0E-07/year. However, this CDF criterion appears to differ from both the 1.0E-06/year CDF criterion noted in Attachment 2 of the Seeping Fire Modeling Scenario Report and the 1.0E-06 conditional core damage probability (CCDP) criterion stated in the disposition to F&O 8-3. Additionally, based on the disposition to F&O 8-3, 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. Describe the process for screening non-propagating fire scenarios, and justify the criteria utilized.

t) F&O 5-24 against FSS-D1 0:

The disposition to this F&O does not address the deficiencies identified by the peer review team and appears to be an editorial error. Provide a disposition to this F&O.

u) F&O 5-26 against FSS-A2:

It is unclear from the disposition to this F&O whether conduits in areas without electrical cabinets, cable trays, or other secondary combustibles were considered by the seeping transient analysis. Given that damage to conduits from a transient fire could result in potential risk-significant scenarios, clarify whether such scenarios were considered by the

seeping transient analysis. If not, provide the results of a sensitivity study (i.e., CDF, LERF, L'lCDF and L'lLERF) that considers conduits.

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 -1 E-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, L'lCDF and L'lLERF) that uses a 1E-5 minimum.

PRA RAI 02- Transient Fire Placement at Pinch Points 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 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.

Per NUREG/CR-6850, Section 11.5.1.6, transient fires should at a minimum be placed in locations within the plant PAUs where CCDPs are highest for that PAU (i.e., at "pinch points").

Pinch points include locations of redundant trains or the vicinity of other potentially risk-relevant equipment. Transient fires should be placed at all appropriate locations in a PAU where they can threaten pinch points. Hot work should be assumed to occur in locations where hot work is a possibility, even if improbable, keeping in mind the same philosophy. Describe how transient and hot work fires are distributed within the PAUs at your plant. In particular, identify the criteria for your plant used to determine where an ignition source is placed within the PAUs.

PRA RAI 03- Transient Fire Influencing Factors 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 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.

A review of Attachment 6 indicates that very high maintenance influencing factor (50) was not applied to any PAUs. Clarify whether influencing factors were developed per the guidance in NUREG/CR 6850 as modified by FAQ 12-0064 (ADAMS Accession No. ML12346A488) and justify why the full range of influencing factors was not implemented.

PRA RAI 04 -Treatment of Alternate Shutdown 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 an FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA 805. In a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing 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. 6 (Control Room Abandonment CCDP and LEAP Calculations) of the Fire Risk Quantification report (i.e., TVA Fire PRA- Task 7.14 Fire Risk Quantification) explains that there are two shutdown paths credited for alternate shutdown (i.e., the 25-32 panel and the emergency high pressure supplemental injection pump) and lists the HFEs credited for each path. Attachment 16 also states that failure of any of these actions is assumed to fail the shutdown path, but does not present failure probability of abandonment used in scenario analysis. Given the complexity of this analysis further describe how the HRA was performed for alternate shutdown following control room abandonment. Include in this description:

a) Identification of fire areas/compartments that credit MCR abandonment due to loss of habitability and/or loss of control or function.

b) Explanation of the logic modeled by the FPRA for crediting the alternate shutdown paths and the basis for that logic (e.g., do the two paths provide redundancy for all fires that result in abandonment scenarios?).

c) Explanation of the range of probabilities for properly shutting down the plant, and discussion of how they were applied in the scenario analysis.

d) Explain how LEAF was determined for abandonment scenarios.

e) Identification of events or conditions that prompt the decision to transfer command-and-control from the MCR. Include clarification of how the decision to abandon will be addressed by the post-transition fire procedures.

f) Explanation of how timing was established (i.e., total time available, time until a cue is reached, manipulation time, and time for decision-making) and which fire or fires were used

as the basis for the timing for both shutdown pathways. Include in the explanation the basis for any assumptions made about timing.

g) Justification for using a cognitive error probability of 0.0 and a Tdelay of 0.0 minutes for alternate-shutdown-related actions as observed in Attachment D of the Fire HRA report (i.e.,

TVA Fire PRA- Task 7.12 Post-Fire Human Reliability Analysis).

h) Description of how the feasibility of the operator actions supporting the two alternate shutdown pathways was assessed. Include a discussion of the potential for detrimental actions.

i) Justification for assuming that continuous communication and coordination will occur during implementation of the alternate shutdown procedure by the different operators at their different locations. Include consideration of actions that require taking off headsets or the unavailability of phone systems.

j) Description of how the impact of complexity on coordination of actions and operator performance in the alternate shutdown procedure addressed, including the complexity that having two alternate shutdown paths injects.

k) Description of the treatment of potential dependencies between individual actions, including discussion of operator actions that can impact the actions of other operators.

I) Discussion of how hardware failure probabilities are addressed (see F&O 7-16).

PRA RAI 05 - Use of Deviations from Accepted Methods and Approaches 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 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.

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. ML12171A583), 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, f1CDF, and f1LERF.

PRA RAI 06 - Fire-Induced Instrument Failure 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 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.

Fire-induced instrument failure should be addressed in the HRA per NUREG/CR-6850 and NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines." Describe how fire-induced instrument failure (including no readings, off-scale readings, and incorrect/misleading readings) is addressed in the fire HRA.

PRA RAI 07- Fire PRA Modeling of HVAC 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 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.

LAR Attachment C identifies a number of systems for which heating ventilation and air conditioning (HVAC) is needed to meet its performance goal. Discuss the HVAC modeling performed to support the FPRA and whether HVAC cable tracing and fire modeling were performed to support this modeling.

PRA RAI 08 - Smoke Damage 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 a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing 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 justification to allow the NRC staff to complete its review of the proposed method.

Explain how the effect of smoke on equipment was evaluated (e.g., by using the guidance provided in Appendix T of NUREG/CR-6850.)

PRA RAI 09 - Sensitive Electronics 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 a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing 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 justification to allow the NRC staff to complete its review of the proposed method.

Describe how sensitive electronics components are identified and treated. Additionally, indicate whether this treatment is consistent with FAQ 13-0004 and the guidance in NUREG/CR-6850.

PRA RAI 10 - Area Wide Incipient Detection 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 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 16-0. 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 for the total flooding system planned for Fire Compartment 16-A.

PRA RAI 11 - Model Changes and Focused Scope Reviews since Full Peer Review 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.

Identify any changes made to the internal events (lEPRA) or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for Nuclear Power Plant Applications," as endorsed by RG 1.200. Also, address the following:

a) Clarify why the SRs in Table V-1 were found to be not applicable (N/A) by the peer review team. Identify which SRs are not applicable (e.g., the type of reactor or the NFPA 805 application). If there are specific and multiple reasons that SRs are determined to be N/A, identify the reason for each SR, grouping them as appropriate. In general, SRs that are not deemed applicable to an application require application specific justification, so all reasons provided should include a discussion of the NFPA 805 application characteristics. For example:

i. FSS-C7 requires that dependencies among multiple suppression paths be evaluated and modeled. Since there are fire scenarios crediting, at a minimum, at least one type of automatic suppression and possible manual suppression if automatic fails, describe the basis for dismissing this SR as inapplicable.

ii. IGN-A6 requires use of Bayesian updating (or equivalent) when generic and plant-specific ignition frequency data are combined. Describe whether any plant-specific datum is used for Browns Ferry at any of the units. If no, why not? If used, describe why this SR is deemed not-applicable.

iii. HRA-B4 addresses the role of instrumentation failures due to fire in HRA. Dismissal of this as inapplicable suggests that such instrumentation failures were not considered in the HRA. Clarify and justify this SR being inapplicable.

iv. For these to be inapplicable (QNS-C and QNS-D2 on p. V-9), QNS-A, -Band -D should also be deemed such, but they do not appear to be.

b) At the time of the follow-on peer review, it is not clear that fire modeling was well developed.

The fire scenario analysis was developed under conservative assumptions for target damage (F&O 3-13), credited for detection and suppression for all target failures without consideration for time to damage or system activation (F&O 3-13), utilized a one-point model (F&O 8-3), discounted "non-propagating" fire scenarios (F&O 5-18), and employed an unreviewed analysis method for severity factors (F&O 3-14). The current fire scenario analysis, however, represents all risk-significant fire scenarios as detailed scenarios that employ a progression of target damage states in time and that credit automatic and manual

suppression at each damage state considering the systems that have been determined to activate at that time. As a result, it appears that the updates made to the fire scenario analysis may be characterized as a PRA upgrade.

Describe whether the above or other changes are considered PRA upgrades. If not, provide justification. If so, indicate if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by RG 1.200, Revision 2, and describe any findings from that focused-scope peer review and the resolution of these findings.

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

PRA RAI12- Short- and Long-Term HFEs 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, L'lCDF and L'lLERF) that applies short- and long-term actions considering the impact of system failures due to both fire-induced and random equipment failures.

PRA RAI 13 - Special Treatment of Emergency Depressurization 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 Post-Fire HRA report states that actions related to emergency depressurization require "special treatment to allow a more refined application in the fire model"; however, the treatment appears to be non-conservative. The discussion in Section 5.1.6 suggests that in an attempt to simplify modeling logic, fire-induced operation of safety relief valves (SRVs) are not modeled as failing high pressure coolant injection (HPCI)/reactor core isolation coolant but are still credited for successful long-term emergency depressurization. Clarify the treatment of actions related to emergency depressurization as well as the modeling of fire-induced operation of SRVs.

PRA RAI 14 - Implementation Item Impact on Risk Estimates 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.

Implementation Items 32 and 33 do not commit to verifying the validity of the reported risk results. When the effect of a plant modification or procedure change has been included in the PRA before the modification or procedural change has been completed, the models and values used in the PRA are necessarily estimates based on current plans. The as-built, as-operated facility after the modification or procedural change is completed may be different than the current estimates. Add an implementation item that verifies the validity of the reported change in risk subsequent to completion of all PRA-credited modifications, procedures updates, and implementation items. This item should include a plan of action should the as-built, as-operated change-in-risk exceed the risk acceptance guidelines.

PRA RAI15- Shorting Switches 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 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 justification to allow the NRC staff to complete its review of the proposed method. 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 ~LEAF) 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.

PRA RAI 16 - Transient Heat Release Rate 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 a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing 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 justification to allow the NRC staff to complete its review of the proposed method.

Section 4.1.4 of the Detailed Fire Modeling Report indicates that reductions in 98th-percentile HRR of NUREG/CR-6850 for transient fires are credited. Identify all fire zones for which a reduction in the HRR below 317 kW for transient fires is credited. In addition, address the location-specific attributes and considerations, plant administrative controls, the results of a review of records related to violations of the transient combustible and hot work controls, and any other key factors used to justify a reduced HRR per the guidance endorsed by the June 21, 2012, memo from Joseph Giitter to Biff Bradley ("Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires"'). If a reduced HRR cannot be justified using these guidance criteria, discuss the impact on the analysis.

PRA RAI 17 - Main Control Room Fire Scenarios 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 a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing 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 justification to allow the NRC staff to complete its review of the proposed method.

Describe how MCR modeling was performed. In doing so:

a) Explain the extent to which propagation of fires between adjacent cabinets in the MCR was evaluated, and provide justification for this level of treatment.

b) Confirm the configurations of ignition sources and secondary targets/combustibles utilized to calculate MCR abandonment times appropriately bound the times for those fire scenarios not explicitly modeled.

c) Provide a quantitative justification that the method utilized to apportion transient fire frequency bounds MCR risk associated with transient fires. Section 8.3 of the MCR analysis indicates that transient fire frequency was arbitrarily apportioned by the number of transient fire scenarios postulated within the MCR. This approach does not appear to address the risk significance of impacted targets (e.g., pinch points) and may underestimate MCR risk.

d) Justify the apportionment of fire frequencies to cable trays within the MCR. Section 9.4 of the MCR analysis indicates that frequency is apportioned by tray length and not weight or combustible loading as suggested by NUREG/CR-6850.

e) Discuss how multi-unit impacts are addressed (e.g., abandonment, fire-induced failures, etc.).

PRA RAI 18 - Calculation of VFDR ~CDF and ~LERF 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.

LAR Attachment W, Section W .2.1 provides description of how the ~CDF and ~LERF for the VFDRs and how the additional risk of RAs for each of the fire areas were determined but does not include any discussion about the use of acceptable methods and no discussion of how the special case of MCR was assessed. The described approach is based on "removing" VFDRs by eliminating fire induced failure and giving perfect credit to RAs associated with VFDRs. It is not clear whether there were any exceptions to this approach, such as in the case of the MCR.

Clarify the following:

a) Clarify how the compliant and variant plants were modeled to determine the ~CDF and

~LERF for the control room and cable spreading room abandonment scenarios.

b) Explain whether there were any exceptions to the approach described in Section W.2.1 (e.g., when there was not enough resolution to the PRA), and describe the approach used for those exceptions.

PRA RAI 19 - Large Reduction Credit for Modifications 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.

LAR Attachment W, Table W-2 presents .b.CDF for each fire area including values that represent exceptionally large risk reductions (e.g., Fire Areas 03-02, 08, 16, and 26). The total .b.CDF for Units 1, 2, and 3 respectively is -5.37E-4/yr, -4.67E-4/yr, and -5.45E-4/yr. Section 3.2.5 of RG 1.205 states that risk decreases may be combined with risk increases for the purposes of evaluating combined changes in accordance with Regulatory Positions 2.1.1 and 2.1.2 of RG 1.174. Accordingly, both individual and cumulative risk effects should be evaluated in detail.

a) Given that the submitted application represents a change that combines risk increases with risk decreases, provide the total increase and total decrease in the .b.CDF and .b.LERF.

b) LAR Attachment W summarizes the risk significant scenarios in the variant case.

Summarize the top few risk significant scenarios for risk significant fire areas in the compliant case.

c) Tables W .8, W .9 and W .1 0 of the LAR show that LERF is generally two to three orders of magnitude less than CDF for most fire areas. Discuss why the fire LEAFs are significantly lower percentage-wise than those for internal events for most fire areas.

PRA RAI 20- Operator Manual Actions (OMAs)

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 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.

PRA RAI 21 - Proposed Plant Modifications 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, LEAF, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates.

LAR Attachment S identifies numerous proposed plant modifications. Although many modifications are extensive most are only briefly described. Provide the following:

a) Summarize how the design of the new features has been provided to the PRA analysts for use in modeling the risk impact (e.g., as brief descriptions or completed design package).

b) Summarize the new models that have been developed (e.g., what basic events, fault trees, event trees, and failure data).

c) Describe how the effect of all new cables has been evaluated (e.g., have areas that credit for the new equipment is being taken been identified as areas where required cables may not be routed).

d) In regard to the disposition to F&O 9-2, which indicates that the single event (i.e., 0.1) representation of the new emergency high-pressure injection pump has been replaced with more detailed modeling logic, describe the scope and content of any interviews, thermal-hydraulic analyses, and procedure reviews performed to support the PRA modeling of this future modification. Additionally, provide a review of modified accident sequences (including timing considerations), system dependencies, flow path considerations, required operator actions, and any new parameter estimates (including unavailability due to testing and maintenance).

PRA RAI22- Circuit Failure Probabilities 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 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 justification to allow the NRC staff to complete its review of the proposed method.

Summarize the circuit analysis method(s) utilized to estimate circuit failure probabilities in the fire PRA (e.g., whether Option 1 and/or 2 in NUREG/CR-6850 were applied). If Option 2 was utilized, clarify how uncertainty was addressed.

PRA RAI 23 - Internal Events PRA F&Os 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. 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 LAR Attachment U that have the potential to impact the FPRA results and do not appear to be fully resolved:

a) 1-12 (Finding against SC-81 Met at CC-II, SY-86 Not Met at CC-1/11/111, and SY-87 Met at CC-1/11/111)

The F&O cites the lack of engineering bases for three specific assumptions; 1) that active ventilation is not required, 2) keep-fill systems are not required, and 3) the high pressure coolant injection (HPCI) system is dependence on "quad" cooling for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of post-accident operation. The disposition to the F&O does not provide a complete basis for these assumptions. Clarify the following:

i. The disposition states that room heat-up calculations were performed after the peer review that resulted in removal of several HVAC system dependencies. Describe the room heat-up calculations that were performed, identify what modeling tools were used, explain which dependencies were removed as a result of these calculations, and identify where this analysis documented.

ii. The disposition explains that the keep fill systems are reliable (i.e., monitored daily by operations and based on operator experience system leaks are not great enough to create a water hammer condition). Describe the engineering basis for not modeling the keep-fill function in the PRA.

iii. The disposition states that based on operational experience and operator interview, the condensate and condensate booster pumps are assumed to survive 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without forced cooling. However, it is noted that Modification 104 identified in Table S-2 of the LAR commits to upgrading the Unit 3 condensate booster pumps to be larger capacity pumps. Clarify whether dependency on forced cooling for these pumps was modeled or removed, and if this dependency was removed, provide justification that the new condensate booster pumps are not dependent on forced cooling.

b) 2-23 (Finding against SY-A3 Met at CC-1/11/111)

Since this F&O was issued modeling of HVAC dependency has been removed for the Unit 3 Electric Board Rooms for both the lEPRA and FPRA. Confirm that HVAC operation is not assumed in the fire PRA to protect equipment from smoke damage or sensitive electronics from thermal damage. Also, confirm that HVAC operation is not needed to ensure access/egress through fire areas during a response.

c) 2-35 (Finding against LE-01 Met at CC-I/II)

Section 3.1.4 of the internal events LERF report (i.e., LE.01 -Probabilistic Risk Assessment- LERF Analysis) presents unit differences including differences in the ability to cross-tie Residual Heat Removal (RHR) and differences in RHR emergency power alignment. These differences are related to ultimate containment structural capability, but appear to impact LERF values, given that RHR is a contributor to In-Vessel Recovery.

Clarify, whether unit differences related to RHR were explicitly modeled in the internal events PRAto determine LERF. If these differences were not incorporated, justify this modeling exclusion or provide an estimate of the impact of this exclusion on internal events and fire PRA CDF and LERF.

d) 2-41 (Finding against LE-C6 Not Met at CC-1/111111)

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.

e) 3-10 (Finding against HR-13 Not Met at CC-1/11/111, IE-03 Not Met at CC-1/11/111, LE-F3 Not Met at CC-1/11/111, SY-C3 Not Met at CC-1/11/111, SC-C3 Not Met at CC-1/11/111, QU-E2 Met at CC-1/11/111, QU-E4 Not Met at CC-1/111111, QU-F4 Not Met at CC-1/11/111, and QU-E1 Not Met at CC-1/11/111)

A detailed compilation of identified sources of uncertainty is presented in Table A8.1 of the internal events Quantification report (i.e., QU - Probabilistic Risk Assessment-Quantification) which includes rationale for assumptions made and indication of internal events PRA modeling impact. However, there appears to be no assessment provided of the potential impacts of these uncertainties to risk estimates. Attachment U of the LAR indicates for this F&O that the fire PRA has its own uncertainty analysis. Explain how the results of the internal events uncertainty analysis were used to inform the fire PRA.

f) 3-31 (Finding against QU-06 Met at CC-1/11/111, QU-F3 Met at CC-1/11/111, QU-F6 Not Met at CC-1/11/111, and LE-G6 Not Met at CC-1/11/111)

Section 6.3.2.3 of the internal events Quantification report (QU - Probabilistic Risk Assessment- Quantification) states that the top 100 CDF and LERF cutsets were considered to define "significance". Explain what contribution the top 100 CDF and LERF cutsets make to the total values, and justify the definition of significance used to define and review significant basic events, cutsets, and accident sequences.

g) 4-4 (Finding against IE-C6 Not Met at CC-1/111111)

This F&O notes that instrument tap failure (i.e., leaks) are screened from the internal events PRA. The F&O disposition explains that instrument lines were considered to contribute to the very small LOCA initiating event category, and are therefore not otherwise addressed.

Explain how the potential for fire-induced failure of instrument lines was addressed in the fire PRA. Include discussion of potential line failure and the effect of impacts of fire on the pressure signal.

h) 4-32 (Findings against SY-A17 Not Met at CC-1/11/111, SY-AB Met at CC-1/11/111, and SY-89 Met at CC-1/11/11 I)

The disposition to this F&O explains that per NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants," external equipment used to monitor under voltage, ground faults, differential faults, and other protection schemes for individual breakers are considered part of the breaker, and therefore failures related to such equipment was assumed to be comprised by the failure of the breaker. Clarify whether the breaker coordination evaluation in the fire PRA would make this F&O inapplicable to the fire PRA. Otherwise, explain how circuit breaker protection monitoring is accounted for in the fire PRA. In particular discuss fire impact on cables associated with the associated protection scheme and interlocks located or extending outside of the cabinet where the circuit breaker is housed that could fail the function of the circuit breaker.

i) 4-45 (Finding against LE-C3 Met at CC-1/11/111)

The disposition to this F&O refers to repair credited in the LEAF model. Explain whether repair is credited in the FPRA and whether fire impacts on repair actions are considered.

j) 5-3 (Finding against DA-82 Not Met at CC-1/11/111)

Section 5.0 of the internal events Data Analysis report (i.e., the DA.01 -Probabilistic Risk Assessment- Data Analysis) states that plant specific raw data was reviewed to identify outlier components and that none were found; however, that review or related data analysis was not found in the report. Describe how raw data was analyzed for outliers and criteria used to determine an outlier.

k) 5-30 (Finding against DA-C13 Not Met at CC-I or /IIIII I)

The disposition to the F&O indicates that the potential for coincident maintenance events was reviewed and additional discussion was added to the internal events Data Analysis report (i.e., DA.01 -Probabilistic Risk Assessment- Data Analysis) to address the impact of differences in Technical Specifications for shared systems of multi-unit plants related plant status. This discussion, however, does not appear to be presented in the cited report.

Explain this inconsistency or describe how the impact of the difference in Technical Specifications for shared systems was addressed when determining test and maintenance unavailability.

I) 6-41 Finding against SY -B 11 Met at CC-1/11/111)

The emergency diesel generators (EDGs) boundary definition from NUREG/CR-6928 cited in the F&O disposition does not provide explicit guidance on whether the diesel fuel oil transfer pump and line from the 7-day diesel storage tank to each diesel generator should be defined to be within the EDG boundary, but it is noted that the fuel oil transfer pumps and lines from the 7-day diesel storage tanks are not physically collocated with the EDGs. This non-proximity could produce fire impacts that must be considered in the fire PRA. Describe how these components are modeled in the FPRA.

Radioactive Release (RR)

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.

RR RAI 02 a) In the Service Building, describe whether there are 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-Radiologial 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.

LICENSEE'S RESPONSE SCHEDULE RAI Number Date and Type of Response Justification (in days)

Fire Protection Engineering (FPE)

FPE 01 1/14/2014 (60)

FPE 02 1/14/2014 (60)

FPE03 1/14/2014 (60)

FPE 04 1/14/2014 (60)

FPE 05 1/14/2014 (60)

FPE 06 1/14/2014 (60)

FPE07 1/14/2014 (60)

FPE 08 1/14/2014 {60)

FPE 09 1/14/2014 (60)

FPE10 2/13/2014 {90) Note 1 FPE 11 2/13/2014 (90) Note 2 FPE12 2/13/2014 (90) Note 2 FPE13 3/15/2014 (120) Note 3 Safe Shutdown Analysis (SSA)

SSA01 1/14/2014 (60)

SSA02 1/14/2014 (60)

SSA03 1/14/2014 (60)

SSA04 1/14/2014 (60)

SSA05 1/14/2014 (60)

Enclosure 2

RAI Number Date and Type of Response Justification (in days)

SSA06 1/14/2014 (60)

SSA07 1/14/2014 (60)

SSAOB 1/14/2014 (60)

SSA09 1/14/2014 (60)

SSA10 1/14/2014 (60)

SSA 11 1/14/2014 (60)

SSA12 1/14/2014 (60)

SSA13 1/14/2014 (60)

SSA14 1/14/2014 (60)

SSA15 1/14/2014 (60)

Programmatic Programmatic 01 1/14/2014 (60)

Programmatic 02 1/14/2014 (60)

Fire Modeling (FM}

FM 01, part a 2/13/2014 (90) Note 4 FM 01, part b.1 1/14/2014 (60)

FM 01, part b.ii 1/14/2014 (60)

FM 01, part b.iii 2/13/2014 (90) Note 5 FM 01, part c 1/14/2014 (60)

FM 01, part d.i 1/14/2014 (60)

FM-01, part d.ii 2/13/2014 (90) Note 6

RAI Number Date and Type of Response Justification (in days)

FM 01, parte 1/14/2014 (60)

FM 01, partf 1/14/2014 (60)

FM 01, part g 2/13/2014 {90) Note 7 FM 01, part h.i 1/14/2014 (60)

FM 01, part h.ii 1/14/2014 (60)

FM 01, part h.iii 2/13/2014 {90) Note 8 FM 01, part i.i 3/15/2014 (120) Note 9 FM 01, part i.ii 1/14/2014 (60)

FM 01, part i.iii 2/13/2014 (90) Note 10 FM 01, part i.iv 3/15/2014 {120) Note 9 FM 01, part i.v 1/14/2014 (60)

FM 01, part i.vi 1/14/2014 (60)

FM 01, part i.vii 2/13/2014 (90) Note 10 FM 01, part i.viii 1/14/2014 {60)

FM 01, part j.i 1/14/2014 (60)

FM 01, part j.ii 2/13/2014 (90) Note 11 FM 02, part a 3/15/2014 {120) Note 12 FM 02, part b 3/15/2014 (120) Note 12 FM 02, part c 1/14/2014 (60)

FM 02, part d 1/14/2014 (60)

FM 02, parte 2/13/2014 (90) Note 12

RAI Number Date and Type of Response Justification (in days)

FM 03 1/14/2014 (60)

FM 04 2/13/2014 (90) Note 13 FM 05 1/14/2014 (60)

FM 06 1/14/2014 (60)

Probabilistic Risk Assessment (PRA}

PRA 01, part a 1/14/2014 (60)

PRA 01, part b 1/14/2014 (60)

PRA 01, part c 2/13/2014 (90) Note 10 PRA 01, part d 2/13/2014 (90) Note 14 PRA 01, part e 3/15/2014 (120) Note 15 PRA 01, part f 3/15/2014 (120) Note 16 PRA 01, part g 1/14/2014 (60)

PRA 01, part h 2/13/2014 (90) Note 16 PRA 01, part i 1/14/2014 (60)

PRA 01, part j 1/14/2014 (60)

PRA 01, part k 1/14/2014 (60)

PRA 01, part I 1/14/2014 (60)

PRA 01, part m 1/14/2014 (60)

PRA 01, part n 1/14/2014 (60)

PRA 01, part o 3/15/2014 (120) Note 16 PRA 01, part p 2/13/2014 (90) Note 17

RAI Number Date and Type of Response Justification (in days)

PRA 01, part q 1/14/2014 (60)

PRA 01, part r 1/14/2014 (60)

PRA 01, parts 2/13/2014 (90) Note 18 PRA 01, part t 1/14/2014 (60)

PRA 01, part u 1/14/2014 (60)

PRA 01, part v 3/15/2014 (120) Note 16 PRA02 1/14/2014 (60)

PRA03 1/14/2014 (60)

PRA04 2/13/2014 (90) Note 19 PRA05 1/14/2014 (60)

PRA06 1/14/2014 (60)

PRA07 1/14/2014 (60)

PRA08 1/14/2014 (60)

PRA09 1/14/2014 (60)

PRA10 2/13/2014 (90) Note 2 PRA 11 1/14/2014 (60)

PRA12 3/15/2014 (120) Note 16 PRA13 1/14/2014 (60)

PRA14 1/14/2014 (60)

PRA15 2/13/2014 (90) Note 16 PRA16 2/13/2014 (90) Note 20

RAI Number Date and Type of Response Justification (in days)

PRA17 2/13/2014 (90} Note 21 PRA18 1/14/2014 (60}

PRA19 1/14/2014 (60)

PRA20 3/15/2014 (120) Note 22 PRA21 1/14/2014 (60)

PRA22 1/14/2014 (60)

PRA23 1/14/2014 (60)

Radioactive Release (RR)

RR 01 1/14/2014 (60)

RR 02 1/14/2014 (60)

NOTES

1. Requires Modification Scoping and a Design Change Document.

2. Requires detailed evaluation and coordination with multiple groups to provide response.

3. May require a re-quantification of CDF/LERF numbers.

4. Requires walkdowns to confirm that non-cable intervening combustibles will not increase damage. Sensitivity analysis and subsequent re-quantification may need to be performed.

5. Requires walkdowns to confirm no holes exist in tray covers. Sensitivity analysis and subsequent re-quantification may need to be performed.

6. Requires walkdowns to review Plexiglass doors in order to confirm approach taken is correct.

Sensitivity analysis and subsequent re-quantification may need to be performed.

7. Requires confirmatory walkdowns of Unit 3 Mechanical Equipment Room to analyze this oil fire scenario. Sensitivity analysis and subsequent re-quantification may need to be performed.

8. Sensitivity analysis needed for non-HGL calculations (i.e., Seyler or FPA).

9. FDS sensitivity analysis required.

10. Requires walkdowns to confirm approach taken is correct. Sensitivity analysis and subsequent re-quantification may need to be performed.

11. CFAST sensitivity analysis required to confirm approach taken is correct.

12. Requires confirmatory walkdowns to confirm approach taken is correct. Sensitivity analysis and subsequent re-quantification may need to be performed.

13. Due to large numbers of scenarios, confirmation of justification of applying models within limits may require additional time.

14. Multiple groups involved in this response. Additional time required to coordinate response effort.

15. Large number of dependencies and HFEs to review. Additional time needed to perform these reviews.

16. Sensitivity study is/may be required to provide response.

17. Bayesian update and possible re-quantification may need to be performed.

18. Quantification of numerous scenarios required to confirm criterion used is correct.

19. RAI has multiple parts and may require re-quantification to ensure approach taken is correct.

20. Response requires a review of records and data collection.

21. Requires quantification justification and potential sensitivity study.

22. RAI has multiple parts and may require a sensitivity study.

ML13298A702 OFFICE LPLII-2/PM LPLII-2/LA APLAIBC AFPB/BC AKiein NAME FSaba BCiayton HHamzehee (JRobinson for)

DATE 11/7/13 11/7/13 11/12/13 11/12/13 OFFICE AHPB/BC LPLII-2/BC LPLII-2/PM FSaba NAME US hoop JQuichocho (Slingam for)

DATE 11/12/13 11/19/13 11/19/13