ML13246A277

Enclosure 3 - Revised Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, Transition Report, August

ML13246A277
Person / Time
Site:	Brunswick
Issue date:	08/28/2013
From:	Carolina Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
BSEP 13-0097, TAC ME9623, TAC ME9624
Download: ML13246A277 (383)

Text

Enclosure 3 Revised Brunswick NFPA 805 Transition Report, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-BasedStandardfor Fire Protection for Light Water ReactorElectric GeneratingPlants, 2001 Edition, Transition Report, August 28, 2013 Main Report Without Attachments

Carolina Power & Light Brunswick Steam Electric Plant Units 1 and 2 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition Transition Report August 28, 2013

CP&L NFPA 805 Transition Report TABLE OF CONTENTS Executive Summary ............................................................................................... iv Acronym List .......................................................................................................... v

1.0 INTRODUCTION

..................................................................................................... 1 1.1 Backg round .................................................................................................. .. 1 1.1.1 NFPA 805 - Requirements and Guidance ................................................. 1 1.1.2 Transition to 10 CFR 50.48(c) ................................................................ 2 1.2 P urpose ...................................................................................................... .. 3 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM ............................ 4 2.1 Current Fire Protection Licensing Basis ......................................................... 4 2.2 NRC Acceptance of the Fire Protection Licensing Basis ............................... 4 3.0 TRANSITION PROCESS .................................................................................... 8 3.1 B ackg ro und .................................................................................................. .. 8 3.2 NFPA 805 Process ........................................................................................ 8 3.3 NEI 04 NFPA 805 Transition Process .................................................. 10 3.4 NFPA 805 Frequently Asked Questions (FAQs) .......................................... 11 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS ........................................ 12 4.1 Fundamental Fire Protection Program and Design Elements ...................... 12 4.1.1 Overview of Evaluation Process ........................................................ 12 4.1.2 Results of the Evaluation Process ...................................................... 14 4.1.3 Definition of Power Block and Plant .................................................... 15 4.2 Nuclear Safety Performance Criteria ........................................................... 15 4.2.1 Nuclear Safety Capability Assessment Methodology ........................... 15 4.2.2 Existing Engineering Equivalency Evaluation Transition .................... 23 4.2.3 Licensing Action Transition .................................................................. 24 4.2.4 Fire Area Transition ............................................................................. 24 4.3 Non-Power Operational Modes .................................................................. 28 4.3.1 Overview of Evaluation Process ......................................................... 28 4.3.2 Results of the Evaluation Process ...................................................... 31 4.4 Radioactive Release Performance Criteria .................................................. 32 4.4.1 Overview of Evaluation Process ........................................................ 32 4.4.2 Results of the Evaluation Process ...................................................... 32 4.5 Fire PRA and Performance-Based Approaches .......................................... 50 4.5.1 Fire PRA Development and Assessment ............................................. 51 I BSEP LAR Rev 2 Page i

CP&L NFPA 805 Transition Report 4.5.2 Performance-Based Approaches ......................................................... 52 4.6 Monitoring Program .................................................................................... 57 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program ...... 57 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ............... 58 4.7 Program Documentation, Configuration Control, and Quality Assurance ........ 63 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 8 0 5 ..................................................................................................... . . 63 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805 ............................................................................... 66 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 .... 70 4.8 Summary of Results ..................................................................................... 72 4.8.1 Results of the Fire Area Review ........................................................ 72 4.8.2 Plant Modifications and Items to be Completed During the Implementation P hase ................................................................................................ . . 73 4.8.3 Supplemental Information -Other Licensee Specific Issues ................ 73

5.0 REGULATORY EVALUATION

......................................................................... 82 5.1 Introduction- 10 CFR 50.48 ....................................................................... 82 5.2 R egulatory Topics ....................................................................................... 87 5.2.1 License Condition Changes ................................................................ 87 5.2.2 Technical Specifications ...................................................................... 87 5.2.3 Orders and Exemptions ...................................................................... 87 5.3 Regulatory Evaluations ................................................................................ 87 5.3.1 No Significant Hazards Consideration ................................................. 87 5.3.2 Environmental Consideration ............................................................. 87 5.4 Revision to the UFSAR .................................................................................. 88 5.5 Transition Implementation Schedule ........................................................... 88

6.0 REFERENCES

.................................................................................................. 89 ATTACHMENTS ....................................................................................................... 91 A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program &

Design Elements ........................................................................................... A-1 B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review ............................................................................................................ B -1 C. NEI 04-02 Table B Fire Area Transition .................................................. C-1 D. NEI 04-02 Non-Power Operational Modes Transition ................................. D-1 E. NEI 04-02 Radioactive Release Transition .................................................. E-1 Page ii Rev 2 BSEP LAR Rev 2 Page ii

CP&L NFPA 805 Transition Report F. Fire-Induced Multiple Spurious Operations Resolution ............................. F-1 G. Recovery Actions Transition ....................................................................... G-1 H. NFPA 805 Frequently Asked Question Summary Table ............................ H-1 I. Definition of Power Block .............................................................................. I-1 J. Fire Modeling V&V ......................................................................................... J-1 K. Existing Licensing Action Transition ........................................................... K-1 L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii) .... L-1 M. License Condition Changes ......................................................................... M-1 N. Technical Specification Changes ............................................................... N-1

0. Orders and Exemptions ................................................................................ 0-1 P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4) ................................... P-1 Q. No Significant Hazards Evaluations ........................................................... Q-1 R. Environmental Considerations Evaluation ................................................. R-1 S. Modifications and Implementation Items .................................................... S-1 T. Clarification of Prior NRC Approvals ........................................................... T-1 U. Internal Events PRA Quality ......................................................................... U-1 V. Fire PRA Q uality ........................................................................................... V-1 W. Fire PRA Insights ........................................................................................ W-1 I Page iii I I BSEP BSEPLARRev2 LAR Rev 2 Page iii

CP&L Executive Summary Executive Summary CP&L will transition the Brunswick Steam Electric Plant (BSEP), Units 1 and 2 fire protection program to a new Risk-Informed, Performance-Based (RI-PB) alternative per 10 CFR 50.48(c) which incorporates by reference NFPA 805. The licensing basis per License Condition 2.B.(6) will be superseded.

The transition process consisted of a review and update of BSEP documentation, including the development of a Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850 as guidance. This Transition Report summarizes the transition process and results. This Transition Report contains information:

" Required by 10 CFR 50.48(c).

" Recommended by guidance document Nuclear Energy Institute (NEI) 04-02 Revision 2 and appropriate Frequently Asked Questions (FAQs).

" Recommended by guidance document Regulatory Guide 1.205, Revision 1.

Section 4 of the Transition Report provides a summary of compliance with the following NFPA 805 requirements:

" Fundamental Fire Protection Program Elements and Minimum Design Requirements

" Nuclear Safety Performance Criteria, including:

o Non-Power Operational Modes o Fire Risk Evaluations o Radioactive Release Performance Criteria

" Monitoring Program

" Program Documentation, Configuration Control, and Quality Assurance Section 5 of the Transition Report provides regulatory evaluations and associated attachments, including:

" Changes to License Condition

" Changes to Technical Specifications, Orders, and Exemptions

" Determination of No Significant Hazards and evaluation of Environmental Considerations The attachments to the Transition Report provide detail to support the transition process and results.

Attachment H contains the approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this License Amendment Request.

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CP&L Acronym List CP&L Acronym List Acronym List ABH Auxiliary Boiler House AC Alternating Current AC/DC Alternating Current/Direct Current ACLE Allowable Combustible Load Equivalent ADAMS Agency wide Documents Access and Management System ADANX Admin - Annex Building (Security Office Building)

ADS Automatic Depressurization System AFEB Alternate Fire Equipment Building AFFF Aqueous Film Forming Foam AHJ Authority having jurisdiction ANS American Nuclear Society AO Auxiliary Operator AOG Augmented Off-Gas Building AOV Air Operated Valve APCSB Auxiliary and Power Conversion Systems Branch ASME American Society of Mechanical Engineers ASSD Alternate Safe Shutdown ASTM American Society for Testing and Materials ATWS Anticipated Transient Without Scram BGA Brunswick Global Analysis BKR Breaker BNP Brunswick Nuclear Plant (i.e., BSEP)

BOP Balance of Plant BSEP Brunswick Steam Electric Plant, Units 1 and 2 BTP Branch Technical Position BWR Boiling Water Reactor BWROG Boiling Water Reactor Owner's Group CAC Containment Atmosphere Control I Page v I IBSEP BSEPLARRev2 LAR Rev 2 Page v

CP&L AcoymLs CP&L Acronym List CAFTA Computer Aided Fault Tree Analysis CAP Corrective Action Program CAS Central Alarm Station CASBCH Caswell Beach CAT Capability Category CB Control Building Cause Based Decision Tree Method/Techniques for CBDTM/THERP Human Error Rate Prediction CBT Computer Based Training CBDTM Cause Based Decision Tree Method CC Capability Category CCI Capability Category I CCDF Conditional Core Damage Frequency CCDP Conditional Core Damage Probability CDF Core Damage Frequency CDM Current Design Method CET Core Exit Thermocouples CFAST Consolidated Model of Fire and Smoke Transport CFD Condensate Filter Demineralizer CFR Code of Federal Regulation CGB Cable Gripping Bushing CLB Chlorination Building CLB Current Licensing Basis CLERP Conditional Large Early Release Probability CLK Nelson Firestop CLTTM Silicone Sealant CM Clean Maintenance CP&L Carolina Power and Light CPT Control Power Transformers CR3 Crystal River Unit 3 Nuclear Power Plant CRD Control Rod Drive CRS Control Room Supervisor CS Core Spray

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CP&L Acronym List CP&L Acronym List CSD Cold Shutdown CSS Core Spray System CST Condensate Storage Tank CSW Conventional Service Water CTPH1 Condensate Transfer Pump House Unit 1 CW Circwater Yard CW Circulating Water System CWOD Circulating Water Ocean Discharge DBA Design Basis Accidents DBD Design Basis Document DC Direct Current DFO Diesel Fuel Oil DG Diesel Generator Building DGB Diesel Generator Building DID Defense-in-Depth DSO Director of Site Operations DWT Demineralized Water Tank EC Engineering Change ECCS Emergency Core Cooling System EDB Equipment Database EDG Emergency Diesel Generator EEE Engineering Equivalency Evaluations EEEE Existing Engineering Equivalency Evaluations EHC Electro-Hydraulic Control EOOS Equipment Out-of-Service EPRI Electric Power Research Institute EQ Environmental Qualification ERFBS Electrical Raceway Fire Barrier Systems ESFAS Engineered Safeguards Actuation Signal EY East Yard FC Fire Compartment F&O Facts and Observations Page vii I IBSEP BSEPLARRev2 LAR Rev 2 Page vii

CP&L Acronym List FAQ Frequently Asked Question FB Fire Brigade FDS Fire Dynamics Simulator FDT Fire Dynamics Tools FHA Fire Hazards Analysis FICF Fire Induced Circuit Failure FIN Fix It Now FMEA Failure Modes and Effects Analysis FP Fire Protection FPIP Fire Protection Initiatives Project FPPM Fire Protection Program Manual FPRA Fire Probabilistic Risk Analysis or Assessment FRE Fire Risk Evaluation FSA Fire Safety Analysis FSAR Final Safety Analysis Report FSSPMD Fire Safe Shutdown Program Manager Database FTL Fault Tree Logic GDC General Design Criterion GL Generic License GPAB Global Plant Analysis Boundary GPM Gallons per Minute HCTL Heat Capacity Temperature Limit HEAF High Energy Arcing Fault HEP Human Error Probabilities HEPA High Efficiency Particulate Air HFE Human Failure Event HGL Hot Gas Layer HLP High/Low Pressure Interface HNP Shearon Harris Nuclear Power Plant HP Health Physics HPCI High Pressure Coolant Injection HPI High Pressure Injection I BSEP LAR Rev 2 Page vii

CP&L Acronym List Human Reliability Analysis technical element from the PRA standard HRA Human Reliability Analysis HRE Higher Risk Evolutions HRR Heat Release Rate HSD Hot Shutdown HSM Horizontal Storage Module HSS High Safety Significance HVAC Heating, Ventilation and Air Conditioning HX Heat Exchanger I&C Instrumentation and Controls IE Initiating Event technical element from PRA standard Internal Flood Scenario Development technical element from the PRA standard Internal Flood Source Identification technical element from the PRA standard IPE Individual Plant Examination ISB ISFSI Storage Building ISFSI Independent Spent Fuel Storage Installation KPI Key Performance Indicator KSF Key Safety Function kV Kilovolt kW Kilowatt LA Licensing Action LAR License Amendment Request LCO Limiting Condition of Operation LDSHD Load Shed PRA model basic event LERF Large Early Release Frequency LFS Limiting Fire Scenario LOCA Loss of Coolant Accident LOOP Loss of Off-site Power LOP Loss of Power LOSP Loss of Off-site Power I

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CP&L Acronym List CP&L Acronym List LPCI Low Pressure Coolant Injection LPI Low Pressure Injection LSS Low Safety Significance MAAP Modular Accident Analysis Program MAF Manual Action Feasibility MBOCA Miscellaneous Buildings - Owner Controlled Area MBPA Miscellaneous Buildings Pre-fire Plans - Protected Area MCA Multiple Compartment Evaluation Approach MCC Motor Control Center MCR Main Control Room MEFS Maximum Expected Fire Scenario MHIF Multiple High Impedance Fault MG Motor Generator MO Motor Operated MOS Maintenance Occupancy and Storage MOV Motor Operated Valve MQH Method of McCaffrey, Quintiere, and Harkleroad MSF Members of the Security Force MSIV Main Steam Isolation Valve MSL Main Steam Line MSO Multiple Spurious Operation MSR Moisture Separator Reheater MUD Make-Up Demineralizer MWT Makeup Water Treatment Building NCR Nuclear Condition Report NDE Non-Destructive Examination NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFPA National Fire Protection Association NFPA 805 National Fire Protection Association Standard 805 NPP Nuclear Power Plant NPO Non-Power Operations I Page x I IBSEP BSEPLARRev2 LAR Rev 2 Page x

CP&L Acronym List CP&L Acronym List NPOPMD Non-Power Operations Program Manager Database NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSCA Nuclear Safety Capability Assessment NSEL Nuclear Safety Equipment List NSP Non-Suppression Probability NSW Nuclear Service Water NUREG US Nuclear Regulatory Commission Regulation NWY Northwest Yard O&M Operations and Maintenance OCA Owner Controlled Area OMA Operator Manual Action OMB Operations/Maintenance Building OOS Out-of-Service ORAM Outage Risk Assessment and Management OSI PI Real Time plant data tracking software OSP On-Site Power PAM Post-Accident Monitoring PB Performance Based PBAA Power Block Auxiliary Areas PDC Power Distribution Center PFP Pre-Fire Plans PGM Plant General Manager P&ID Piping and Instrumentation Diagram PLC Professional Loss Control PMP Pump PNL Panel PNSC Plant Nuclear Safety Committee POM Plant Operating Manual PORV Power Operated Relief Valves POS Plant Operational State I Page xi I I BSEP BSEPLARRev2 LAR Rev 2 Page xi

CP&L Acronym List PRA Probabilistic Risk Assessment or Analysis PVC Polyvinyl-chloride PWR Pressurized Water Reactor Fire Qualitative Screening technical element from the PRA standard QU Quantification technical element from the PRA standard RA Recovery Actions RAI Request for Additional Information RAW Risk Achievement Worth RB Reactor Building RBCCW Reactor Building Closed Cooling Water RCA Radiologically Controlled Area RCIC Reactor Core Isolation Cooling RCR Reactor Coolant Recirculation RCS Reactor Coolant System RFP Reactor Feed Pump RFPT Reactor Feed Pump Turbine RG Regulatory Guide RHR Residual Heat Removal RI-PB Risk-Informed Performance-Based RIS Regulatory Issues Summary RMA Radioactive Materials Area RMCSB Radioactive Material - Container Storage Building RPDC Recirc Power Distribution Center RPS Reactor Protection System RPV Reactor Pressure Vessel RSDP Remote Shutdown Panel RW Radwaste RWB Radwaste Building RWCU Reactor Water Cleanup System SAMA Severe Accident Mitigation Alternative SAP Secondary Access Point

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CP&L AcovmLs CP&L Acronym List SAR Safety Analysis Report SAT Startup Auxiliary Transformer SBGT Standby Gas Treatment SBO Station Blackout SCAFF Clean Scaffold Material Storage SCBA Self Contained Breathing Apparatus SD System Description SDC Shutdown Cooling SDV Scram Discharge Volume SE Safety Evaluation SER Safety Evaluation Report SFPC Spent Fuel Pool Cooling SFPE Society of Fire Protection Engineers SHF Sodium Hypochlorite Facility SIC Site Incident Commander SJAE Steam Jet Air Ejector SLC Standby Liquid Control SM Safety Margin SP Suppression Pool SPC Suppression Pool Cooling SR Supporting Requirement SRV Safety Relief Valve SSA Safe Shutdown Analysis SSC Structures, Systems, and Components SSD Safe Shutdown SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List STORES Hot Shop/Material Issue/Warehouse STORM Storm Drain Monitoring SW Service Water SWB Service Water Building SWGR Switchgear

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CP&L Acronym List CP&L Acronym List SWY Switchyard SY Switchyard TAP Training Administrative Procedure TB Turbine Building TS Technical Specification UAT Unit Auxiliary Transformer UFSAR Updated Final Safety Analysis Report VFDR Variances from the deterministic requirements VFDs Variable Frequency Drives V&V Verification and Validation WFSS Water-based Fire Suppression System WW Wet well ZOI Zone of Influence I Page xiv I IBSEP BSEPLARRev2 LAR Rev 2 Page xiv

CP&L 1.0 Introduction

1.0 INTRODUCTION

The Nuclear Regulatory Commission (NRC) has promulgated an alternative rule for fire protection requirements at nuclear power plants, 10 CFR 50.48(c), National Fire Protection Association Standard 805 (NFPA 805). CP&L is implementing the Nuclear Energy Institute methodology NEI 04-02, "Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program Under 10 CFR 50.48(c)", to transition BSEP from its current fire protection licensing basis to the new requirements as outlined in NFPA 805. This report describes the transition methodology utilized and documents how BSEP complies with the new requirements.

1.1 Background 1.1.1 NFPA 805 - Requirements and Guidance On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes new Risk-Informed, Performance-Based (RI-PB) fire protection requirements. 10 CFR 50.48(c) incorporates by reference, with exceptions, the National Fire Protection Association's NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition, as a voluntary alternative to 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.

As stated in 10 CFR 50.48(c)(3)(i), any licensee's adoption of a RI-PB program that complies with the rule is voluntary. This rule may be adopted as an acceptable alternative method for complying with either 10 CFR 50.48(b), for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979, or 10 CFR 50.48(f), plants shutdown in accordance with 10 CFR 50.82(a)(1).

NEI developed NEI 04-02 to assist licensees in adopting NFPA 805 and making the transition from their current fire protection licensing basis to one based on NFPA 805.

The NRC issued Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants, which endorses NEI 04-02, with exceptions, in December 2009.1 A depiction of the primary document relationships is shown in Figure 1-1:

1Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.

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CP&L 1.0 CP&L 1.0 Introduction Introduction NFPA 805 Incorporation by 50.48(c) 2001 ed. Reference Performance-Based g Standard for FP for National Fire Light Water Reactor Protection Electric Generating Association Plants Standard NFPA 805 NEI 04-02 RG 1.205 EI Endorsement OGUIDANCE FOR ' RI-PB FP FOR EXISTING IMPLEMENTING A RI-PB LIGHT-WATER NUCLEAR FP PROGRAM UNDER 10 POWER PLANTS CFR 50.48(c)

Figure 1-1 NFPA 805 Transition - Implementation Requirements/Guidance 1.1.2 Transition to 10 CFR 50.48(c) 1.1.2.1 Start of Transition CP&L submitted a letter of intent to the NRC on June 10, 2005 (ML051720404), for the Shearon Harris Nuclear Power Plant (HNP) to adopt NFPA 805 in accordance with 10 CFR 50.48(c). This letter of intent also addressed other CP&L plants (Brunswick Steam Electric Plant Units No. 1 and 2, H.B. Robinson Steam Electric Plant Unit No. 2, and Crystal River Unit 3 Nuclear Generating Plant). The letter of intent requested three years of enforcement discretion and proposed that HNP be considered a Pilot Plant for the NFPA 805 transition process.

By letter dated April 29, 2007 (ML070590625), the NRC granted a three year enforcement discretion period. In accordance with NRC Enforcement Policy, the enforcement discretion period will continue until the NRC approval of the license amendment request (LAR) is completed.

1.1.2.2 Transition Process The transition to NFPA 805 includes the following high level activities:

0 Complete Safe Shutdown Analysis Reconstitution (activities started in 2003)

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CP&L 1.0 Introduction

" A new Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, as guidance and a revision to the Internal Events PRAs to support the Fire PRAs

" Completion of activities required to transition the pre-transition licensing basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205 The project was implemented using a comprehensive project plan and individual procedures/instructions for individual scopes of work. These procedures/instructions (e.g., Project Instruction "FPIP" series procedures referenced in this report) were developed for the purposes of NFPA 805 transition. Appropriate technical content from these procedures were and will be incorporated into technical documents and configuration control procedures, as required.

1.2 Purpose The purpose of the Transition Report is as follows:

1) Describe the process implemented to transition the current fire protection program to comply with the additional requirements of 10 CFR 50.48(c).

2) Summarize the results of the transition process.

3) Explain the bases for conclusions that the fire protection program complies with 10 CFR 50.48(c) requirements.

4) Describe the new fire protection licensing basis.

5) Describe the configuration management processes used to manage post-transition changes to the station and the fire protection program, and resulting impact on the licensing basis.

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CP&L 2.0 Overview of Existing Fire Protection Program 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM 2.1 Current Fire Protection Licensing Basis Brunswick Steam Electric Plant was licensed to operate on September 8, 1976, for Unit 1 and December 27, 1974, for Unit 2. As a result, the Brunswick Steam Electric Plant fire protection program is based on evaluation and NRC acceptance against the requirements of Design Criterion 3, Appendix A to 10 CFR 50 Part 50, and 10 CFR 50 Appendix R, Sections III.G and J. The following License Condition 2.B(6) in Amendment No. 169 to the Facility Operating License No. DPR-71 (i.e., Unit 1) and Amendment No. 200 to Facility Operating License No. DPR-62 (i.e., Unit 2) states:

"CarolinaPower and Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report, dated November 22, 1977, as supplemented April 1979, June 11, 1980, December 30, 1986, December 6, 1989, and July 28, 1993 and February10, 1994, respectively, subject to the following provision:

The licensee may make changes to the approved fire protection program without priorapproval of the Commission only if those changes would not adversely affect the ability to achieve and maintainsafe shutdown in the event of a fire."

2.2 NRC Acceptance of the Fire Protection Licensing Basis The Commission issued, on November 22, 1977, Amendment No. 11 to the Facility Operating License No. DPR-71, for Unit No. 1, and Amendment No. 37 to Facility Operating License No. DPR-62, for Unit No. 2, of the Brunswick Steam Electric Plant.

These amendments added license conditions relating to the completion of the facility modifications for fire protection and resolution of incomplete items. The amendment for Unit 1 also incorporated limiting conditions for operations and surveillance requirements for existing fire protection systems and administrative controls.

Amendment No. 11 contained the following changes to 2.B(6) and 2.C.(2):

2.B(6) The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.35 of the NRC's Fire Protection Safety Evaluation Report on the Brunswick facility dated November 22, 1977.

These modifications shall be completed by the end of the first refueling outage of Brunswick Unit 1 and prior to return to operation of Cycle 2. In addition, the licensee shall submit the additional information identified in Table 3.1 of this Safety Evaluation Report in accordance with the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report explaining the circumstances, together with a revised schedule.

2.C.(2) The Technical Specifications contained in Appendices A, A-Prime and B, attached hereto, as revised through Amendment No. 11, are hereby incorporated in this license. Appendix A shall be effective from the date of I

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CP&L 2.0 Overview of Existing Fire Protection Program issuance of the Unit 1 operating license until the Appendix A-Prime becomes effective on or before the initial criticality of Brunswick Unit 2 following its initial refueling outage. Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications as indicated above.

The licensee shall inform the Office of Inspection and Enforcement, Region II, of the date that the Appendix A-Prime becomes effective.

Amendment No. 37 contained the following changes to 2.B(7):

2.B(7) The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.35 of the NRC's Fire Protection Safety Evaluation Report on the Brunswick facility dated November 22, 1977.

These modifications shall be completed by the end of the second refueling outage of Brunswick Unit 2 and prior to return to operation of Cycle 3. In addition, the licensee shall submit the additional information identified in Table 3.1 of this Safety Evaluation Report in accordance with the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report explaining the circumstances, together with a revised schedule.

The Commission issued, on April 6, 1979, Amendment No. 23 to the Facility Operating License No. DPR-71, for Unit No. 1, and Amendment No. 47 to Facility Operating License No. DPR-62, for Unit No. 2, of the Brunswick Steam Electric Plant. These amendments consisted of changes to the operating licenses for both units to allow revised implementation dates for certain modifications intended to improve the level of fire protection. Supplement 1 of the Fire Protection Safety Evaluation Report was also included in this transmittal which addressed certain items that were identified as incomplete and requiring further information from the licensee and evaluation by the staff. The SER, Supplement 1, also listed several modifications proposed by the licensee to improve fire protection.

The Commission issued Supplement 2 to the Fire Protection Safety Evaluation Report on June 11,1980, which contained evaluations associated with four areas: 1) Protection of Redundant Safe Shutdown Cabling (greater than five foot separation, 2) Protection of Redundant Safe Shutdown Cabling (less than five foot separation), 3) Fire Protection Loop Isolation Valve, 4) Door Frames for Fire Doors.

The Commission granted, on November 10, 1981, an exemption from the requirements of 10 CFR 50 Appendix R, Section III.G.3, with regard to fixed fire suppression in the Control Room.

The Commission granted, on July 27, 1983, exemptions from the requirements of 10 CFR 50 Appendix R, Section III.G.3, with regard to fixed fire suppression in the seven fire zones in the Control Building Cable Vaults.

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CP&L 2.0 Overview of Existing Fire Protection Program The Commission granted, on September 17, 1986, an exemption from the licensee commitment to install an excess flow switch and automatic shut-off valve in the fuel supply line for the diesel fire pump to automatically isolate in the event of a fuel line rupture. The Safety Evaluation concluded that the previous commitment to provide automatic isolation of the diesel fuel line need not be implemented because of the alternative fire protection measures provided.

The Commission granted, on December 30, 1986, exemptions from the requirements of Appendix R to 10 CFR Part 50, Sections IIL.G and J. Exemptions were granted for the following:

1) Reactor Buildings, Units 1 and 2 (Fire Areas RBI-1 and RB2-1)

2) Emergency Core Cooling System Rooms, Units 1 and 2 (Fire Areas RB1-6 and RB2-6)

3) Diesel Generator Building Basement (Fire Area DG-1)

4) Service Water Building (Fire Area SW-I)

5) Diesel Generator Building (DG-08)

6) Fixed Fire Suppression System for Alternative Shutdown Areas (Fire Areas TB-i, CB-1, CB-7, CB-8, CB-9, CB-10, DG-6, DG-7, DG-9, DG-11, DG-12, DG-13 and DG-14)

7) East Yard Area The Staff concluded that the exemption request for the Control Building Extended (Fire Area CB-23E) was not needed.

The Commission issued, on May 29, 1987, a Safety Evaluation approving the use of higher unexposed side temperatures for fire barrier seals than that required by the Branch Technical Position (BTP) ASB 9.5-1 of NUREG-0800. The evaluation concluded that the acceptance criteria of 325 *F above ambient, versus 250 'F above ambient, was an acceptable deviation and was not considered likely to significantly add to the risk of igniting material on the unexposed side of the barrier.

The Commission granted, on August 27, 1987, an Exemption from 10 CFR Part 50, Appendix R, Section Ill.J, from the requirement for emergency lighting units with at least an 8-hour battery supply in all areas needed for operation of safe shutdown equipment.

The exemption permits substitution of 8-hour battery lighting with:

1) The use of diesel generators to power lighting in the plant control room upon loss of offsite power.

2) The use of two-hour battery-powered lighting upon loss of diesel generators concurrent with loss of offsite power.

3) Assurance that power sources are routed underground and are separated by at least a three-hour rated fire barrier.

The Commission issued an Appendix R Safety Evaluation Clarification and Revision on December 6, 1989. Brunswick Steam Electric Plant had identified nineteen items I

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CP&L 2.0 Overview of Existing Fire Protection Program associated with the Staff's December 30, 1986, Safety Evaluation where revisions were required to 1) correct specific errors, 2) clarify potentially confusing language, or 3) more accurately state actual conditions. The Staff provided clarifications for fifteen of the nineteen items requested. These clarifications appended the December 30, 1986, Safety Evaluation.

The Commission issued, on July 28, 1993, a Safety Evaluation approving a request to downgrade the three-hour rated masonry block walls in the control building cable access ways (separating fire areas CB-01a/b, CB-02a/b, CB-12a/b and CB-13a/b) to non-rated walls. The Staff found this change did not have an adverse impact on the III.G.3 exemption granted for the lack of fire suppression in the control building and would not impact the alternate shutdown capability.

The Commission issued, on February 10, 1994, a Safety Evaluation that revised the plant fire protection licensing condition and Technical Specifications (TS). In accordance with Generic Letter 86-10 and 88-12, CP&L requested that fire protection be removed from the Technical Specifications and a standard fire protection licensing condition be implemented. The following Technical Specification changes were proposed and granted by the NRC:

1) Delete TS 3.3.5.7 (Fire Detection Instrumentation), TS 3.7.7.1 (Fire Suppression Water System), TS 3.7.7.2 (Spray and/or Sprinkler Systems), TS 3.7.7.3, (High Pressure Carbon Dioxide), TS 3.7.7.4 (Fire Hose Stations), TS 3.7.7.5, (Foam Systems), and TS 3.7.8 (Fire Barrier Penetrations) and their associated bases and incorporate into the Updated Final Safety Analysis Report (UFSAR).

2) Delete TS 6.2.2.g for site fire brigade staffing and incorporate into the UFSAR

3) Delete TS 6.4.2 requirements related to the fire brigade training program and incorporate into the UFSAR.

4) Add TS 6.5.3.8(m) to include the review of the fire protection program and implementing procedures as an additional responsibility of the Plant Nuclear Safety Committee (PNSC).

5) Delete TS 6.9.2.d related to the requirement for special reports for the fire detection instrumentation.

6) Delete TS 6.9.2.g related to the requirement for special reports for fire suppression systems.

7) Delete TS 6.9.2.h related to the requirement for special reports for fire barrier penetrations.

CP&L also proposed, and the NRC granted the request, to replace the current fire protection licensee condition with the standard license condition provided in GL 86-10.

I Page 7 I IBSEP BSEPLARRev2 LAR Rev 2 Page 7

CP&L 3.0 Transition Process 3.0 TRANSITION PROCESS 3.1 Background Section 4.0 of NEI 04-02 describes the process for transitioning from compliance with the current fire protection licensing basis to the new requirements of 10 CFR 50.48(c).

NEI 04-02 contains the following steps:

1) Licensee determination to transition the licensing basis and devote the necessary resources to it;

2) Submit a Letter of Intent to the NRC stating the licensee's intention to transition the licensing basis in accordance with a tentative schedule;

3) Conduct the transition process to determine the extent to which the current fire protection licensing basis supports compliance with the new requirements and the extent to which additional analyses, plant and program changes, and alternative methods and analytical approaches are needed;

4) Submit a LAR;

5) Complete transition activities that can be completed prior to the receipt of the License Amendment;

6) Receive a Safety Evaluation; and

7) Complete implementation of the new licensing basis, including completion of modifications identified in Attachment S.

3.2 NFPA 805 Process Section 2.2 of NFPA 805 establishes the general process for demonstrating compliance with NFPA 805. This process is illustrated in Figure 3-1. It shows that, except for the fundamental fire protection requirements, compliance can be achieved on a fire area basis either by deterministic or RI-PB methods. Consistent with the guidance in NEI 04-02, CP&L has implemented the NFPA 805, Section 2.2 process, by first determining the extent to which its current fire protection program supports findings of deterministic compliance with the requirements in NFPA 805. RI-PB methods are being applied to the requirements for which deterministic compliance could not be shown.

I Page 8 I BSEPLARRev2 BSEP LAR Rev 2 Page 8

CP&L 3.0 Transition Process Establish fundamental fire NFPA 805 Section 2.2(a) protection elements (Chapter 3)

Identify fire hazards NFPA 805 Section 2.2(b)

I, Nuclear safety Life safety NFPA 805 Section 2.2(c) Property damage/business Interruption Evaluate compliance to Radiation release performance criteria NFPA 805 Section 2.2(d)

NFPA 805 Section 2.2(e)

DeterministicApproach Performance-BasedApproach Maintain compliance with existing plant Evaluate ability to satisfy performance I* -._

license basis (10 CFR 50 App. R, Approved requirements Exemptions, Engineering Evaluations)

(Chapter 4)

[

DeterministicBasis ExitngI Engineering PerformanceBasis Define fire scenarios and fire design basis Verify deterministic requirements are met for each fire area being considered.

Evaluations Evaluate using, e.g.,

" Fire modeling to quantify the fire risk NFPA 805 Section 2.2(f) and margin of safety

" PSA to examine impact on overall plant risk NFPA 805 Section 2.2(g)

Risk-Informed Change Evaluation NFPA 805 Section 2.2(h)

Evaluate risk impact of changes to the approved design basis acetbe No Examples Yes Design Basis Documents I t* Fire hazards analysis NFPA 805 Section 2.20) Documentation and configuration control Nuclear safety Supporting capability calculations engineering assessment

• , Probabilistic safety analysis Risk-informed change evaluations Establish monitoring program NFPA 805 Section 2.2(i)

Figure 3-1 NFPA 805 Process [NEI 04-02 Figure 3-1 based on Figure 2-2 of NFPA 805]2 2 Note: 10 CFR 50.48(c) does not incorporate by reference Life Safety and Plant Damage/Business Interruption goals, objectives and criteria. See 10 CFR 50.48(c) for specific exceptions to the incorporation by reference of NFPA 805.

I I IBSEP LAR Rev 2 Page 9

CP&L 3.0 Transition Process 3.3 NEI 04 NFPA 805 Transition Process NFPA 805 contains technical processes and requirements for a RI-PB fire protection program. NEI 04-02 was developed to provide guidance on the overall process (i.e.,

programmatic, technical, and licensing) for transitioning from a traditional fire protection licensing basis to a new RI-PB method based upon NFPA 805, as shown in Figure 3-2.

Section 4.0 of NEI 04-02 describes the detailed process for assessing a fire protection program for compliance with NFPA 805, as shown in Figure 3-2.

Transition Report Transition Report Sect. 4.1 Sect. 4.2 FP Fundamentals Review Nuclear Safety Review and Confirmation and Confirmation Transition Report Transition Report Identify outliers / VFDRs Identify outliers / VFDRs Sect. 4.4 Sect. 4.3 r

Perform Engineering Analyses Non-power FP Radioactive operational Nuclear Safety Fundamentals Release mode Analyses Assessment Assessment Assessment I.

Use PB Approach if Needed (Fire Modeling or Report Fire Risk Evaluations) Sect 4.5 Transition Verify / Establish Monitoring Report Program

+Transition Sect. 4.6 Transition Confirm / Establish Adequate Documentation / Quality and Report Configuration Control Sect. 4.7, 5

                                                                                 }  Transition Regulatory Submittal and Approval                           Report Sect 4.8, 5 Figure 3-2 Transition Process (Simplified) [based on NEI 04-02 Figure 4-1]

I Page 10 IBSEPLARRev2 I BSEP LAR Rev 2 Page 10

CP&L 3.0 Transition Process 3.4 NFPA 805 Frequently Asked Questions (FAQs) The NRC has worked with NEI and two Pilot Plants (Oconee Nuclear Station and HNP) to define the licensing process for transitioning to a new licensing basis under 10 CFR 50.48(c) and NFPA 805. Both the NRC and the industry recognized the need for additional clarifications to the guidance provided in RG 1.205, NEI 04-02, and the requirements of NFPA 805. The NFPA 805 FAQ process was jointly developed by NEI and NRC to facilitate timely clarifications of NRC positions. This process is described in a letter from the NRC dated July 12, 2006, to NEI (ML061660105) and in Regulatory Issue Summary (RIS) 2007-19, Process for Communicating Clarifications of Staff Positions Provided in RG 1.205 Concerning Issues Identified during the Pilot Application of NFPA Standard 805, dated August 20, 2007 (ML071590227). Under the FAQ Process, transition issues are submitted to the NEI NFPA 805 Task Force for review, and subsequently presented to the NRC during public FAQ meetings. Once the NEI NFPA 805 Task Force and NRC reach agreement, the NRC issues a memorandum to indicate that the FAQ is acceptable. NEI 04-02 will be revised to incorporate the approved FAQs. This is an on-going revision process that will continue through the transition of NFPA 805 plants. Final closure of the FAQs will occur when future revisions of RG 1.205, endorsing the related revisions of NEI 04-02, are approved by the NRC. It is expected that additional FAQs will be written, and existing FAQs will be revised, as plants continue NFPA 805 transition after the Pilot Plant Safety Evaluations. Attachment H contains the list of approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this LAR. I Page 11 I IBSEP BSEPLARRev2 LAR Rev 2 Page 11

CP&L 4.0 Compliance with NFPA 805 Requirements 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS 4.1 Fundamental Fire Protection Program and Design Elements The Fundamental Fire Protection Program and Design Elements are established in Chapter 3 of NFPA 805. Section 4.3.1 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis and plant configuration meets these criteria and for identifying the fire protection program changes that would be necessary for compliance with NFPA 805. NEI 04-02 Appendix B-1 provides guidance on documenting compliance with the program requirements of NFPA 805 Chapter 3. 4.1.1 Overview of Evaluation Process The comparison of the BSEP Fire Protection Program to the requirements of NFPA 805 Chapter 3 was performed and documented in Attachment A, Table B-i, NFPA 805 Ch. 3 Transition Details. The analysis used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 (See Figure 4-1). Each section and subsection of NFPA 805 Chapter 3 was reviewed against the current fire protection program. Upon completion of the activities associated with the review, the following compliance statement(s) was used:

       " Complies - For those sections/subsections determined to meet the specific requirements of NFPA 805.
       " Complies with Clarification - For those sections/subsections determined to meet the requirements of NFPA 805 with clarification.
       " Complies by previous NRC approval - For those sections/subsections where the specific NFPA 805 Chapter 3 requirements are not met but previous NRC approval of the configuration exists.
       " Complies with use of Existing Engineering Equivalency Evaluations (EEEEs) -

For those sections/subsections determined to be equivalent to the NFPA 805 Chapter 3 requirements as documented by engineering analysis.

       " Submit for NRC Approval - For those sections/subsections for which approval is sought in this LAR submittal in accordance with 10 CFR 50.48(c)(2)(vii). A summary of the bases of acceptability is provided (see Attachment L for details).

In some cases, multiple compliance statements have been assigned to a specific NFPA 805 Chapter 3 section/subsection. Where this is the case, each compliance/compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3. I Page 12 I IBSEP BSEPLARRev2 LAR Rev 2 Page 12

CP&L 4.0 Compliance with NFPA 805 Requirements progromand esign Element Existin Fundamenta Fiirewrotoýon tnterpFoela provide: ieDocument Refereneh that Engi'nrgq Ev atinNooeoy additional damonstrae ompliance Domnnt anyOpen the NFPA805 Chapter 3 Requirement Clarification'

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Reurd nCmpineSaeequ ntd licenseamendment *t.-n I found duringReview on transitio the of EEEsistincludgiederin Secton. 4.2.2. BSE LR tev 2 Pae1 Figure 4 Fundamental Fire Protection Program and Design Elements Transition Process [Based on NEI 04-02 Figure 4-2] 3 3 Figure 4-1 depicts the process used during the transition and therefore contains elements (i.e., open items) that represent interim resolutions. Additional detail on the transition of EEEEs is included in Section 4.2.2. BSEP LAR Rev 2 Page 13

CP&L 4.0 Compliance with NFPA 805 Requirements 4.1.2 Results of the Evaluation Process 4.1.2.1 NFPA 805 Chapter 3 Requirements Met or Previously Approved by the NRC Attachment A contains the NEI 04-02, Table B-i, Transition of Fundamental Fire Protection Program and Design Elements. This table provides the compliance basis for the requirements in NFPA 805 Chapter 3. Except as identified in Section 4.1.2.3, Attachment A demonstrates that the fire protection program at BSEP either:

      " Complies directly with the requirements of NFPA 805 Chapter 3,
      " Complies with clarification with the requirements of NFPA 805 Chapter 3,
      " Complies through the use of existing engineering equivalency evaluations which are valid and of appropriate quality, or
      " Complies with a previously NRC approved alternative to NFPA 805 Chapter 3 and, therefore, the specific requirement of NFPA 805 Chapter 3 is supplanted.

4.1.2.2 NFPA 805 Chapter 3 Requirements Requiring Clarification of Prior NRC Approval NPFA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases, prior NRC approval of an NFPA 805 Chapter 3 program attribute may be unclear. CP&L requests that the NRC concur with their finding of prior approval for the following sections of NFPA 805 Chapter 3: 0 None. 4.1.2.3 NFPA 805 Chapter 3 Requirements Not Previously Approved by NRC The following sections of NFPA 805, Chapter 3, are not specifically met nor do previous NRC approvals of alternatives exist:

• 3.5.16 - Approval is requested for the use of fire protection water for specific plant evolutions.
• 3.2.3(1) - Approval is requested for the use of performance-based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805.

The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements are provided in Attachment L. CP&L requests NRC approval of these performance-based methods. Page 14 I IBSEP LAR Rev 2 BSEPLARRev2 Page 14

CP&L 4.0 Compliance with NFPA 805 Requirements 4.1.3 Definition of Power Block and Plant Where used in NFPA 805, Chapter 3, the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility's pre-transition licensing basis. All structures within the BSEP Owner Controlled Area were reviewed to determine the potential impact of fire on the nuclear safety and radioactive release criteria described in Section 1.5 of NFPA 805. This was accomplished by identifying the structures that contain either

            - Equipment that could affect
                   " Plant operation for power generation
                   " Ability to maintain nuclear safety performance criteria in the event of a fire, including Safe Shutdown Capability OR
            - Radioactive materials that could potentially be released in event of a fire These structures are listed in Attachment I and define the "power block" and "plant".

4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805. Chapter 4 of NFPA 805 provides the methodology to determine the fire protection systems and features required to achieve the performance criteria outlined in Section 1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis meets these criteria and for identifying any necessary fire protection program changes. NEI 04-02, Appendix B-2, provides guidance on documenting the transition of Nuclear Safety Capability Assessment Methodology and the Fire Area compliance strategies. 4.2.1 Nuclear Safety Capability Assessment Methodology The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes:

       " Establishing compliance with NFPA 805 Section 2.4.2
       " Establishing the Safe and Stable Conditions for the Plant
       " Establishing Recovery Actions
       " Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3.

I Page 15 I I BSEP BSEPLARRev2 LAR Rev 2 Page 15

CP&L 4.0 Compliance with NFPA 805 Requirements 4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states:

     "The purpose of this section is to define the methodology for performing a nuclearsafety capabilityassessment. The following steps shall be performed:

(1) Selection of systems and equipment and their interrelationshipsnecessaryto achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclearsafety performance criteriain Chapter 1 (3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclearsafety performance criteria given a fire in each fire area" The NSCA methodology review evaluated the NSCA methodology against the guidance provided in NEI 00-01, Revision 1 (ML050310295) Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology is depicted in Figure 4-2 and consisted of the following activities:

     " Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station.
     " The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section:

o Aligns o Aligns with intent o Not in Alignment o Not in Alignment, but Prior NRC Approval o Not in Alignment, but no adverse consequences

     " For those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01 could have adverse consequences. Since NEI 00-01 is a guidance document, portions of its text could be interpreted as 'good practice' or intended as an example of an efficient means of performing the analyses. If the section has no adverse consequences, these sections of NEI 00-01 can be dispositioned without further review.

The comparison of the BSEP existing post-fire Safe Shutdown Analysis (SSA) methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in Attachment B, Table B-2 Nuclear Safety Capability Assessment Methodology Review. In addition, a review of NEI 00-01, Revision 2, (ML091770265) Chapter 3, was conducted to identify the substantive changes from NEI 00-01, Revision 1 that are applicable to an NFPA 805 fire protection program. This review was performed and I BSEP LAR Rev 2 Page 16

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements documented in Attachment B, Table B-2 Nuclear Safety Capability Assessment Methodology Review. Results from Evaluation Process The method used to perform the existing post-fire SSA with respect to selection of systems and equipment, selection of cables, and identification of the location of equipment and cables, either meets the NRC endorsed guidance from NEI 00-01, Revision 1, Chapter 3 (i.e., as supplemented by the gap analysis) directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B. Step I Assemble Documentation r Determine and Document Step 2 Applicability of NEI 00-01 Sections For Applicable NEI 00-01 Sections, Perform Comparison of SSD Method vs. NEI 00-01 Step3 d-< consequences. Step 4

                                                                                                           )

K-Figure 4 Summary of Nuclear Safety Methodology Review Process (FAQ 07-0039)

 !                                                                                                 Page 17 I BSEPLARRev2 BSEP LAR Rev 2                                                                                 Page 17

CP&L 4.0 Compliance with NFPA 805 Requirements Comparison to NEI 00-01 Revision 2 An additional review was performed of NEI 00-01, Revision 2, Chapter 3, for specific substantive changes in the guidance from NEI 00-01, Revision 1 that are applicable to an NFPA 805 transition. The results of this review are summarized below:

     " Post fire manual operation of rising stem valves in the fire area of concern (NEI 00-01 Section 3.2.1.2)

A review of the NSCA results indicated that there are no recovery actions or defense-in-depth recovery actions that require manual operation of a rising stem valve in the fire area of concern.

     " Analysis of open circuits on a high voltage (e.g., 4.16 kV) ammeter current transformers (NEI 00-01 Section 3.5.2.1)

The evaluation concludes that this failure mode is unlikely for CTs that could pose a threat to safe shutdown equipment.

     " Analysis of control power for switchgear with respect to breaker coordination (NEI 00-01 Section 3.5.2.4)

Control power is modeled in the safe shutdown fault tree used to develop the NSCA. A loss of control power results in an assumed loss of the switchgear, and there are no cases where a bus is credited to remain operable without control power. 4.2.1.2 Safe and Stable Conditions for the Plant Overview of Process The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic programs based on 10 CFR 50 Appendix R and NUREG 0800, Section 9.5-1 (and NEI 00-01, Chapter 3) since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown. NFPA 805, Section 1.6.56, defines Safe and Stable Conditions as follows "Forfuel in the reactorvessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactorcoolant temperature at or below the requirements for hot shutdown for a boiling water reactorand hot standby for a pressurizedwater reactor.For all other configurations,safe and stable conditionsare defined as maintainingKeff <0.99 and fuel coolant temperature below boiling." The nuclear safety goal of NFPA 805 requires "...reasonableassurance that a fire during any operationalmode and plant configuration will not prevent the plant from achieving and maintainingthe fuel in a safe and stable condition" without a specific reference to a mission time or event coping duration. For the plant to be in a safe and stable condition, it may not be necessary to perform a transition to cold shutdown as currently required under 10 CFR 50, Appendix R. Therefore, the unit may remain at or below the temperature defined by a hot standby/hot shutdown plant operating state for the event. ! I BSEP LAR Rev 2 Page 18

CP&L 4.0 Compliance with NFPA 805 Requirements Results Based on the criteria discussed in NCSA calculation BNP-E-9.010, "Safe Shutdown Analysis in Case of Fire," the NFPA 805 licensing basis for BSEP is to achieve and maintain hot shutdown conditions following any fire occurring prior to establishing cold shutdown. Specifically, the conditions include:

• the reactor operating at power,
• a shutdown immediately prior to aligning the RHR system for shutdown cooling, or

      " the "transition" mode between these two operational phases.

Immediately following the reactor scram, RCS inventory and pressure control is maintained using the high pressure systems, HPCI and RCIC, or the low pressure injection systems using the SRV's for pressure reduction, which are the RHR System in LPCI mode or Core Spray System. For the most limiting fire scenarios in every fire area, BNP-E-9.010 documents the availability of long term cooling using the RHR system, in either the Normal Shutdown Cooling Mode or Alternate Shutdown Cooling Mode, or the Core Spray System, all of which are characterized by low pressure injection and at least 1 SRV available to provide core flow. The RHR Service Water system rejects decay heat to the ultimate heat sink. Notably, initiation of RHR in the suppression pool cooling mode does not imply that the plant would proceed all the way to cold shutdown. Following stabilization at hot shutdown, a long term strategy for decay heat removal and inventory/pressure control would be determined based on the extent of equipment damage. If an assessment of the post-fire conditions indicated that placing RHR in the Shutdown Cooling or Alternate Shutdown Cooling modes would be advisable, then activities would commence in a safe and controlled manner to align plant equipment required for reactor cooldown. The long-term actions required to maintain safe and stable conditions are relatively low risk activities that are largely routine and within the normal capabilities of site personnel, even in the face of fire damage, due to the assured availability of at least one train of RHR and either onsite or offsite power sources. Repairs to safe shutdown equipment would not be required and the management of the onsite inventories of makeup water, nitrogen and diesel fuel would not require resources beyond those available from normal operations staff and emergency response personnel. Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses.

      > At-Power analysis, Modes 1-3. This analysis is discussed in Section 4.2.4.
      > Non-Power Operations analysis that includes cold shutdown and below, or Modes 4 and 5. This analysis is discussed in Section 4.3.

I Page 19 I IBSEP BSEPLARRev2 LAR Rev 2 Page 19

CP&L 4.0 Compliance with NFPA 805 Requirements 4.2.1.3 Establishing Recovery Actions Overview of Process NEI 04-02 and RG 1.205 suggest that a licensee submit a summary of its approach for addressing the transition of OMAs as recovery actions in the LAR (i.e., Regulatory Position 2.2.1 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02 suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee's methodology. The discussion below provides the methodology used to transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML110070485) and consists of the following steps:

      " Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) Activities that occur in the Main Control Room are not considered pre-transition OMAs. Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition.
      " Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (i.e., to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).
      " Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path
      " Step 4: Evaluate the feasibility of the recovery actions
      " Step 5: Evaluate the reliability of the recovery actions Results The review results are documented in the Fire Safety Analysis for each area. Refer to Attachment G for the detailed evaluation process and summary of the results from the process.

4.2.1.4 Evaluation of Multiple Spurious Operations Overview of Process NEI 04-02 suggests that a licensee submit a summary of its approach for addressing potential fire-induced MSOs for NRC review and approval. As a minimum, NEI 04-02 recommends that the summary contain sufficient information relevant to methods, tools, and acceptance criteria used to enable the NRC to determine the acceptability of the licensee's methodology. The methodology used to address MSOs for Brunswick is summarized below. As part of the NFPA 805 transition project, a review and evaluation of Brunswick susceptibility to fire-induced MSOs was performed. The process was conducted in accordance with NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038 Revision 3 (ML110140242). The BWR Generic MSO list from NEI 00-01, Revision 3 was utilized. ! Page 20 I IBSEP LAR Rev 2 BSEPLARRev2 Page 20

CP&L 4.0 Compliance with NFPA 805 Requirements The approach outlined in Figure 4-3, below, (i.e., based on Figure 4-8 from FAQ 07-0038) is the method used to address fire-induced MSOs for BSEP. This method used insights from the Fire PRA developed in support of transition to NFPA 805 and consists of the following:

      " Identifying potential MSOs of concern.
      " Conducting an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1 Section F.4.2).
      " Updating the Fire PRA model and existing post-fire SSA / NSCA to include the MSOs of concern.
      " Evaluating for NFPA 805 compliance.
      " Documenting results.

This process is intended to support the transition to a new licensing basis. Post-transition changes would use the RI-PB change process. The post-transition change process for the assessment of a specific MSO would be a simplified version of this process, and may not need the level of detail shown in the following section (e.g., an expert panel may not be necessary to identify and assess a new potential MSO. Identification of new potential MSOs will be part of the plant change review process and/or inspection process). I Page 21 I IBSEP BSEPLARRev2 LAR Rev 2 Page 21

CP&L. 4.0 Compliance with NFPA 805 Requirements Identify Potential MSOs of Concern

                            "      SSA
                            "      Generic List of MSOs Step 1
                            "      Self Assessments
                            "      PRA Insights
                            "      Operating Experience
                                                +

Expert Panel Step 2 Identify and Document MSOs of Concern Update PRA model & NSCA (as appropriate) to include MSOs of concern Step 3 " ID equipment

• ID logical relationships
• ID cables
• ID cable routing Compliance Pursue other resolution options Step 4 Yes Step 5 Document Results Figure 4 Multiple Spurious Operations - Transition Resolution Process (Based on FAQ 07-0038)

Results Refer to Attachment F for the process used by BSEP. I Page 22 I IBSEP LAR Rev 2 BSEPLARRev2 Page 22

CP&L 4.0 Compliance with NFPA 805 Requirements 4.2.2 Existing Engineering Equivalency Evaluation Transition Overview of Evaluation Process The EEEEs that support compliance with NFPA 805 Chapter 3 or Chapter 4 (i.e., both those that existed prior to the transition and those that were created during the transition) were reviewed using the methodology contained in NEI 04-02. The methodology for performing the EEEE review included the following determinations:

       " The EEEE is not based solely on quantitative risk evaluations,
       " The EEEE is an appropriate use of an engineering equivalency evaluation,
       " The EEEE is of appropriate quality,
       " The standard license condition is met,
       " The EEEE is technically adequate,
       " The EEEE reflects the plant as-built condition, and
       " The basis for acceptability of the EEEE remains valid In accordance with the guidance in RG 1.205, Regulatory Position 2.3.2 and NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are summarized in the LAR as follows:
       " If not requesting specific approval for "adequate for the hazard" EEEEs, then the EEEE was referenced where required and a brief description of the evaluated condition was provided.
       " If requesting specific NRC approval for "adequate for the hazard" EEEEs, then EEEE was referenced where required to demonstrate compliance and was included in Attachment L for NRC review and approval.

In all cases, the reliance on EEEEs to demonstrate compliance with NFPA 805 requirements was documented in the LAR. Results The review results for EEEEs are documented in Attachment A. In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in the Attachments A and C as appropriate. None of the transitioning EEEEs require NRC approval. ! Page 23 I I BSEP LAR Rev 2 BSEPLARRev2 Page 23

CP&L 4.0 Compliance with NFPA 805 Requirements 4.2.3 Licensing Action Transition Overview of Evaluation Process The existing licensing actions (i.e., Appendix R exemptions) review was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following:

      " Determination of the bases for acceptability of the licensing action.
      " Determination that these bases for acceptability are still valid and required for NFPA 805.
      " Additionally, variances from the deterministic requirements were identified in the NEI 04-02, Table B-3 (See Attachment C). Some of these variances were subsequently dispositioned via the use of the performance-based approach.

Results Attachment K contains the detailed results of the Licensing Action Review. None of the licensing actions will be transitioned into the NFPA 805 fire protection program. The licensing actions listed in Attachment K are no longer necessary and will not be transitioned into the NFPA 805 fire protection program. The justifications, grouped by the nature of the exemption, are provided in Attachment 0, Orders and Exemptions. Since the exemptions are either compliant with 10 CFR 50.48(c) or no longer necessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), CP&L requests that the exemptions listed in Attachment K be rescinded as part of the LAR process. It is CP&L's understanding that implicit in the superseding of the current license condition, all prior fire protection program Safety Evaluations and commitments will be superseded in their entirety. 4.2.4 Fire Area Transition Overview of Evaluation Process The Fire Area Transition (i.e., NEI 04-02 Table B-3) was performed using the methodology contained in NEI 04-02 and FAQ 07-0054. The methodology for performing the Fire Area Transition, depicted in Figure 4-4, is outlined as follows: Step 1 - Assembled documentation. Gathered industry and plant-specific fire area analyses and licensing basis documents. Step 2 - Documented fulfillment of nuclear safety performance criteria.

      " Assessed accomplishment of nuclear safety performance goals. Documented the method of accomplishment, in summary level form, for the fire area.
      " Documented evaluation of effects of fire suppression activities. Documented the evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria.
      " Performed licensing action reviews. Performed a review of the licensing aspects of the selected fire area and documented the results of the review. See Section 4.2.3.

I I BSEP LAR Rev 2 Page 24

CP&L 4.0 Compliance with NFPA 805 Requirements

       "   Performed existing engineering equivalency evaluation reviews. Performed a review of existing engineering equivalency evaluations, or created new evaluations, documenting the basis for acceptability. See Section 4.2.2.
       " Pre-transition OMA reviews. Performed a review of pre-transition OMAs to determine those actions taking place outside of the main control room or outside of the primary control station(s). See Section 4.2.1.3.

Step 3 - VFDR Identification and characterization and resolution considerations. Identified variances from the deterministic requirements of NFPA 805, Section 4.2.3. Documented variances as either a separation issue or a degraded fire protection system or feature. Developed VFDR problem statements to support resolution. Step 4 - Performance-Based evaluations (i.e., Fire Modeling or Fire Risk Evaluations) See Section 4.5.2 for additional information. Step 5 - Final Disposition.

       " Documented final disposition of the VFDRs in the fire safety analysis for each area.
       " For recovery action compliance strategies, ensured the manual action feasibility analysis of the required recovery actions was completed. Note: If a recovery action cannot meet the feasibility requirements established per NEI 04-02, then alternate means of compliance was considered.
       " Documented the post transition NFPA 805 Chapter 4 compliance basis.

Step 6 - Documented required fire protection systems and features. Reviewed the NFPA 805, Section 4.2.3, compliance strategies (i.e., including fire area licensing actions and engineering evaluations) and the NFPA 805, Section 4.2.4, compliance strategies (i.e., including simplifying deterministic assumptions) to determine the scope of fire protection systems and features 'required' by NFPA 805 Chapter 4. The 'required' fire protection systems and features are subject to the applicable requirements of NFPA 805, Chapter 3. I Page 25 I IBSEP LAR Rev 2 BSEPLARRev2 Page 25

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements Figure 4 Summary of Fire Area Review [Based on FAQ 07-0054 Revision 1] I BSEP LAR Rev 2 Page 26

CP&L 4.0 Compliance with NFPA 805 Requirements Results of the Evaluation Process Attachment C contains the results of the Fire Area Transition review (i.e., NEI 04-02, Table B-3). On a fire area basis, Attachment C summarizes compliance with Chapter 4 of NFPA 805. NEI 04-02, Table B-3, includes the following summary level information for each fire area:

      " Regulatory Basis - NFPA 805 post-transition regulatory bases are included.
      " Performance Goal Summary - An overview of the method of accomplishment of each of the performance criteria in NFPA 805 Section 1.5 is provided.
      " Reference Documents - Specific references to Nuclear Safety Capability Assessment Documents are provided.
      " Fire Suppression Activities Effect on Nuclear Safety Performance Criteria - A summary of the method of accomplishment is provided.
      " Licensing Actions - BSEP is not transitioning any existing Licensing Actions, as noted in Attachment K.
      " EEEE - Specific references to EEEE that rely on determinations of "adequate for the hazard" that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability should be provided.
      " VFDRs - Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-based approach.

I Page 27 I IBSEP BSEPLARRev2 LAR Rev 2 Page 27

CP&L 4.0 Compliance with NFPA 805 Requirements 4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process BSEP implemented the process outlined in NEI 04-02, Guidance for implementing a Risk-Informed, Performance-Based Program under 10 CFR 50.48(c), and FAQ 07-0040, Clarification on Non-Power Operations. The goal (i.e., as depicted in Figure 4-6) is to ensure that contingency plans are established when the plant is in a Non-Power Operational (NPO) mode where the risk is intrinsically high. During low risk periods, normal risk management controls and fire prevention/protection processes and procedures will be utilized. The process to demonstrate that the nuclear safety performance criteria are met during NPO modes involved the following steps:

      " Reviewed the existing Outage Management Processes
      " Identified Equipment/Cables:

o Reviewed plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and o Identified cables required for the selected components and determined their routing.

      " Performed Fire Area Assessments (identify pinch points - plant locations where a single fire may damage all success paths of a KSF).
      " Manage pinch-points associated with fire-induced vulnerabilities during the outage.

The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2. ! Page 28 I IBSEP LAR Rev 2 BSEPLARRev2 Page 28

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points I Page 29 I IBSEP LAR Rev 2 BSEPLARRev2 Page 29

CP&L 4.0 Compliance with NFPA 805 Requirements Higher Risk Evolution as Defined by Plant Specific Outage Risk Criteria for example

1) Time to Boil

2) Reactor Coolant System and Fuel Pool Inventory

3) Decay Heat Removal Figure 4-6 Manage Pinch Points I Page 30 IBSEPLARRev2 I BSEP LAR Rev 2 Page 30

CP&L 4.0 Compliance with NFPA 806 Requirements 4.3.2 Results of the Evaluation Process BSEP outage management processes were reviewed. Based on FAQ 07-0040, the Plant Operating States considered for equipment and cable selection are documented in calculation BNP-E-9.01 1, "NFPA 805 Transition - NPO Modes Review." Using a CAFTA fault tree that models NPO requirements, systems and components were identified to provide three KSFs: Decay Heat Removal, Inventory Control, and Electrical Power Availability (i.e., to the extent that it supports the Decay Heat Removal and Inventory Control functions). For those components not already in the BSEP Access Database or those with a functional state for non-power operations differing from that in the At-Power Analysis, circuit analysis, cable selection and routing were performed as described in the plant's NSCA methodology. Once all information had been entered into the BSEP Access Database, the ARCTM software package in conjunction with the NPO fault tree was used to determine KSF Pinch Points. Calculation BNP-E-9.01 1 provides the results of the fire area assessments for the Pinch Point analysis and provides recommendations for changes to fire risk and outage management procedures and other administrative controls. These include:

      " Prohibition or limitation of hot work in fire areas during periods of increased vulnerability.
      " Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability.

• Provision of additional fire watches in affected fire areas during increased vulnerability.
• Identification and monitoring of in-situ ignition sources for "fire precursors" (e.g.,

equipment temperatures).

      " Review of work activities for possible rescheduling
      " Equipment realignment (e.g., Swing pumps, Backfeed, etc.)
      " Identified procedures to be briefed or walked down.
      " Posting of protected equipment.
      " Use of recovery actions to mitigate potential losses of KSF success paths.

Attachment D provides a more detailed discussion. Based on incorporation of the recommendations from BNP-E-9.011 into appropriate plant procedures in conjunction with establishment of the NFPA 805 fire protection program, the performance goal for NPO modes (i.e., maintain KSF availability) is fulfilled and the requirements of NFPA 805 are met. I Page 31 I IBSEP LAR Rev 2 BSEPLARRev2 Page 31

CP&L 4.0 Compliance with NFPA 805 Requirements 4.4 Radioactive Release Performance Criteria 4.4.1 Overview of Evaluation Process The review of the fire protection program against NFPA 805 requirements for fire suppression related radioactive release was performed using the methodology contained in NEI 04-02, Table E-1, and was performed using the methodology contained in Project Instruction FPIP-0121, Radiological Release Reviews During Fire Fighting Operations, Rev. 1. The methodology consisted of the following:

     " A review of fire pre-plans and fire brigade training materials to identify fire protection program elements (e.g., systems / components / procedural control actions / flow paths) that are being credited to meet the radioactive release goals, objectives, and performance criteria during all plant operating modes, including full power and non-power conditions. Specifically for BSEP, a review was conducted by a review panel to ensure specific steps are included for containment and monitoring of potentially contaminated materials so as to limit the potential for release of radioactive materials due to firefighting operations.

The review panel consisted of representatives from Operations, Engineering (i.e., Fire Protection, HVAC Systems), Operations Fire Brigade Training, and Radiation Protection. Site pre-fire plans were screened to identify those locations that have the potential for radiological contamination based on location within plant Radiological Controlled Areas, areas containing potentially contaminated systems, or locations where radioactive materials are routinely stored. In addition, the site fire brigade training materials were reviewed by the same review panel to ensure specific steps are included addressing containment and monitoring of potentially contaminated materials and monitoring of potentially contaminated fire suppression products following a fire event.

     " A review of engineering controls to ensure containment of gaseous and liquid effluents (i.e., smoke and fire fighting agents). This review included all plant operating modes (i.e., including full power and non-power conditions).

Otherwise, provided a bounding analysis, quantitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of radioactive effluents specified in the unit's Technical Specifications are met. 4.4.2 Results of the Evaluation Process Fire Pre-Plan review; The review determined the Fire Protection Program (i.e., Pre-Fire Plans) meets the radioactive release performance criteria by ensuring that radioactive materials (i.e., radiation) generated as a direct result of fire suppression activities is contained and monitored prior to release to unrestricted areas, such that release would be as low as reasonably achievable and would not exceed applicable 10 CFR, Part 20 limits. Containment and monitoring is ensured through elements of the fire brigade training, guidance provided in pre-fire plans and certain plant features (i.e., engineering controls) such as curbs and ventilation systems or actions provided to control smoke management or fire suppression water run-off. I BSEP LAR Rev 2 Page 32

CP&L 4.0 Compliance with NFPA 805 Requirements Site specific review of associated fire event and fire suppression related radioactive release is summarized in Attachment E, NEI 04-02, Table E-1. Containment and monitoring actions associated with fire fighting operations are included in the pre-fire plans for fire areas as appropriate based on the screening criteria previously stated (i.e., ref. Table 4-3 and Attachment E) to meet the radiological performance criteria. The standardized pre-fire plan outline identifies typical fixed radiological hazards for each area. All BSEP pre-fire plans were screened for applicability. Pre-fire plans that address areas where there is no possibility of radiological hazards were screened out from further review. A summary cross-reference of fire compartment, fire area, and pre-fire plan to plant fire areas, and radioactive release input results is provided in Table 4-3. This information was included as input to the individual fire area Fire Safety Analyses (FSA's) calculations. The FSA is the Design Basis Document for NFPA 805 compliance for each fire area and will serve as the location for maintenance and configuration control of the radioactive release review results. Change, modification, or revision to the FSA's is controlled under existing plant engineering configuration control processes. ! Page 33 I IBSEP BSEPLARRev2 LAR Rev 2 Page 33

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements Table 4-3 -BSEP Pre-Fire Plan Screening Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) Y/N Unit 1 Cable Access Way (North East Rattle GB-0lA Unit 1 Cable Access OPFP-CB 1PEP-GB-i Out Y Space) 23ft. Elevation GBiWays _____ ______ ____ Unit 1 Gable Access Way (North West Rattle CB-1 CB-01 B Unit 1 Gable Access OPFP-GB i PFP-CB-1 Out Y Space) 23ft. Elevation Ways Unit 1 Cable Access Way (North East Rattle GB-i CB-12A Unit 1 Cable Access OPFP-CB lPFP-CB-12 Out Y Space) 49ft. Elevation Ways Unit 1 Cable Access Way (North West Rattle GB-i CB-12B Unit 1 Cable Access OPFP-CB iPFP-CB-12 Out Y Space) 49ft. Elevation Ways Battery Room 2B 23ft. Elevation GB-iO GB-10 Unit 2 Division 11Battery OPFP-GB 2PFP-CB-i0 Out Y Room Unit 2 Cable Access Way (South East Rattle CB-2 CB-02A Unit 2 Cable Access OPFP-CB 2PFP-CB-2 Out Y Space) 23ft. Elevation Ways Unit 2 Gable Access Way (South West GB-2 CB-02B Unit 2 Gable Access OPFP-CB 2PFP-CB-2 Out Y Rattle Space) 23ft. Elevation WaysCB-10CB-10Ro Unit 2 Cable Access Way (South East Rattle GB-2 CB-13A Unit 2 Cable Access OPFP-GB 2PFP-CB-13 Out Y Space) 49ft. Elevation Ways Unit 2 Cable Access Way (South West Rattle GB-2 CB-13B Unit 2 Cable Access OPFP-GB 2PFP-GB-13 Out Y Space) 49ft. Elevation Ways Unit 1 Northwest Stairwell 23ft. and 49ft CB-23E CB-03 Gontrol Room Extended OPFP-CB i PFP-CB-4 Out Y Elevations 49tlvainWy Unit 2 Southwest Stairwell 23ft. and 49ft CB-23E CB-04 Gontrol Room Extended OPFP-CB 2PFP-CB-3 Out Y Elevations 49tlvainWy Unit 1 Gable Spreading Room 23ft. CB-23E CB-05 Control Room Extended OPFP-CB 1PFP-CB-5 Out Y Elevation Unit 2 Gable Spreading Room 23ft. CB-23E CB-06 Control Room Extended OPFP-CB 2PFP-CB-6 Out Y Elevation Gontrol Building Elevator and Shaft CB-23E CB-il Control Room Extended OPFP-CB OPFP-CB-1 1 Out Y Unit 1 Computer Room North 49ft. Elevation CB-23E CB-14 Control Room Extended OPFP-CB OPFP-CB-14 Out Y Unit 2 Computer Room South 49ft. Elevation CB-23E CB-15 Control Room Extended OPFP-CB OPFP-CB-14 Out Y I Page 34 I I BSEP LAR Rev 2 BSEPLARRev2 Page 34

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan PrePlan Input RCA? Procedure Pr-PanInu (Screened) Auxiliary Operator Briefing Room CB-23E CB-16 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Operator Break Room CB-23E CB-17 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Ladies Washroom CB-23E CB-18 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Central Alarm Station (CAS) 52ft. Elevation CB-23E CB-19 Control Room Extended OPFP-CB OPFP-CB-19 Out Y Unit 1 Northwest Back Panel Area CB-23E CB-20 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Unit 2 Southwest Back Panel Area CB-23E CB-21 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Men's Washroom CB-23E CB-22 Control Room Extended OPFP-CB OPFP-CB-23 Out Y Control Room 49ft. Elevation CB-23E CB-23 Control Room Extended OPFP-CB OPFP-CB-23 Out Y HVAC Equipment Room - 70ft CB-23E CB-24 Control Room Extended OPFP-CB OPFP-CB-24 Out Y Air Conditioning Condenser Area - 70ft CB-23E CB-25 Control Room Extended OPFP-CB OPFP-CB-25 Out Y Control Building Elevator Machinery Room - CB-23E CB-26 Control Room Extended OPFP-CB OPFP-CB-26 Out Y 70ft CB-23E C-6otlRmEed 0P-B PP -2 Ot Battery Room 1A 23ft. Elevation CB-7 CB-07 Unit Room1 Division I Battery 0PFP-CB 1PFP-CB-7 Out Y Battery Room 1 B 23ft. Elevation CB-8 CB-08 Unit 1 Division II Battery OPFP-CB 1PFP-CB-8 Out Y Battery__Room_1B_23ft._ElevationCB-8_CB-08_Room Battery Room 2A 23ft. Elevation CB-9 CB-09 Unit 2 Division I Battery 0PFP-CB 2PFP-CB-9 Out Y Room DG Building Basement - 2ft DG-1 DG-01 Diesel Generator OPFP-DG OPFP-DG-1 Out N __________Basement DG Building Loading Dock, 20ft DG-10 DG-10 Loading Dock OPFP-DG OPFP-013 Out N El Switchgear Room, 50ft DG-1 1 DG-1 1 El Switchgear OPFP-DG 1PFP-DG-1 1 Out N E2 Switchgear Room, 50ft DG-12 DG-12 E2 Switchgear OPFP-DG 1PFP-DG-12 Out N E3 Switchgear Room, 50ft DG-13 DG-13 E3 Switchgear OPFP-DG 2PFP-DG-13 Out N E4 Switchgear Room, 50ft DG-14 DG-14 E4 Switchgear OPFP-DG 2PFP-DG-14 Out N Supply Air Plenum, 50ft DG-16E DG-15 Fan Room Extended OPFP-DG OPFP-DG-15 Out N Diesel Building Supply Fan Room, 50ft DG-16E DG-16 Fan Room Extended OPFP-DG OPFP-DG-15 Out N Diesel Building North Air Lock, 50ft DG-16E DG-17 Fan Room Extended OPFP-DG OPFP-DG-15 Out N I Page 35 I I BSEP LAR Rev 2 BSEPLARRev2 Page 35

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) YIN Diesel Building South Air Lock, 50ft DG-16E DG-18 Fan Room Extended OPFP-DG OPFP-DG-15 Out N AFFF System Room, 50ft DG-16E DG-23 Fan Room Extended OPFP-DG OPFP-DG-15 Out N Diesel Generator Fuel Oil Tank Cell #1, 2ft DG-19 DG-19 Fuel Oil Tank Cell 1 OPFP-DG 0PFP-DG-19 Out N Diesel Generator Cell 4 - 20ft DG-2 DG-02 Diesel Cell 4 OPFP-DG 2PFP-DG-2 Out N Diesel Generator Fuel Oil Tank Cell #2, 2ft DG-20 DG-20 Fuel Oil Tank Cell 2 OPFP-DG OPFP-DG-19 Out N Diesel Generator Fuel Oil Tank Cell #3, 2ft DG-21 DG-21 Fuel Oil Tank Cell 3 OPFP-DG OPFP-DG-19 Out N Diesel Generator Fuel Oil Tank Cell #4, 2ft DG-22 DG-22 Fuel Oil Tank Cell 4 OPFP-DG OPFP-DG-19 Out N Diesel Generator Cell 3 - 20ft DG-3 DG-03 Diesel Cell 3 OPFP-DG 2PFP-DG-3 Out N Diesel Generator Cell 2 - 20ft DG-4 DG-04 Diesel Cell 2 OPFP-DG 1 PFP-DG-4 Out N Diesel Generator Cell 1, 20ft DG-5 DG-05 Diesel Cell 1 OPFP-DG 1 PFP-DG-5 Out N E5 Switchgear Room, 20ft DG-6 DG-06 E5 Switchgear OPFP-DG 1PFP-DG-6 Out N E6 Switchgear Room, 20ft DG-7 DG-07 E6 Switchgear OPFP-DG 1PFP-DG-7 Out N E7 Switchgear Room, 20ft DG-8 DG-08 E7 Switchgear OPFP-DG 2PFP-DG-8 Out N E8 Switchgear Room, 20ft DG-9 DG-09 E8 Switchgear OPFP-DG 2PFP-DG-9 Out N Caswell Beach Pumping Station CASBCH CASBCH Caswell Beach Pumping OPFP- 0PFP-CAS Out N Station MBOCA Hydrogen/Oxygen Storage Facility HOSF HOSE Hydrogen/Oxygen OPFP-HyrgnOyetrgeFcltOF HS Storage Facility MBOCA 0PFP-HOSF Out N Sodium Hypochlorite Facility SHF SHF Sodium Hypochlorite OPFP-Facility MBOCA 0PFP-SHF Out N Switchyard YARD SY Yard OPFPC 0PFP-RELAY Out N _________ _________MBOCA OF-EA u Auxiliary Boiler House ABH ABH Auxiliary Boiler House OPFP-MBPA OPFP-ABH Out N Admin - Annex Building (Security Office ADANX ADANX Admin - Annex Building OPFP-MBPA OPFP-ADANX Out N Building) (Security Office Building) 0F-BA PPADN OuN Admin Building - First Floor ADMIN ADMIN-01 Administration Building OPFP-MBPA OPFP-ADMIN-1 Out N Admin Building - Second Floor ADMIN ADMIN-02 Administration Building OPFP-MBPA OPFP-ADMIN-2 Out N Clean Maintenance Shop - First Floor CM CM-01 Clean Maintenance Shop OPFP-MBPA OPFP-CM-1 Out N I Page 36 BSEPLARRev2 I BSEP LAR Rev 2 Page 36

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Procedure Pre Pre-Plan Input Input RCA? YIN (Screened) Clean Maintenance Shop - Second Floor CM CM-02 Clean Maintenance Shop OPFP-MBPA OPFP-CM-2 Out N Fab Shop #1/MOV and Electrical FAB FAB-01 Fab Shops OPFP-MBPA OPFP-FAB1 Out N Fab Shop #3/Diesel Repair FAB FAB-03 Fab Shops OPFP-MBPA OPFP-FAB3 Out N Fire House FH FH Fire House OPFP-MBPA OPFP-FH Out N ISFSI Storage Building ISB ISB ISFSI Storage Building OPFP-MBPA OPFP-ISFSI-SB IN Y Lube Oil and Paint Storage Building LUBE LUBE Lube Oil and Paint OPFP-MBPA OPFP-LUBE Out N __________________________________ __________Storage Building ___________ I&C Breaker Test/NDE Building MAINT MAINT l&C Breaker Test0NDE OPFP-MBPA OPFP-MAINT Out N Building ________ Mini Warehouse/Equipment- Outage Mini Warehouse/ OPFP-MBPA MINI MINI Equipment - Outage OPFP-MINI Out N Storage Building Storage Building Makeup Water Treatment Building MWT-1 MWT-01 Makeup Water Treatment OPFP-MBPA 0PFP-MWT Out N O&M Building - First Floor OMB OMB-01 O&M Building OPFP-MBPA OPFP-OMB-1 Out N O&M Building - Second Floor OMB OMB-02 O&M Building OPFP-MBPA OPFP-OMB-2 Out N O&M Building - Third Floor OMB OMB-03 O&M Building OPFP-MBPA OPFP-OMB-3 Out N Radioactive Material - Radioactive Material - Container Storage RMCSB RMCSB Container Storage OPFP-MBPA OPFP-RMCSB IN Y Building __Building Secondary Access Point SAP SAP Secondary Access Point OPFP-MBPA OPFP-SAP Out N Clean Scaffold Building SCAFF SCAFF Clean Scaffold Building OPFP-MBPA OPFP-SCAFF Out N Service Building - First Floor SERV SERV-01 Service Building OPFP-MBPA 0PFP-SERV-1 IN Y Service Building - Second Floor SERV SERV-02 Service Building OPFP-MBPA OPFP-SERV-2 IN Y Hot Shop/Stores/Warehouse Building STORES STORES Hot Shop/Stores/ 0PFP-MBPA 0PFP-STORES IN Y Warehouse Building Storm Drain Rad Monitor Building STORM STORM Storm Drain Rad Monitor OPFP-MBPA OPFP-STORM IN Y Building Augmented Off-Gas Building AOG-1 AOG-01 Augmented Building Off-Gas 0F-BA0F-O1 IN I Page 37 I I BSEP LAR Rev 2 BSEPLARRev2 Page 37

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Procedure Pre Pre-Plan Input Input RCA? Y/N (Screened) Unit 1 Recirc Power Distribution Center RPDC1 RPDC1 Unit 1 Recirc Power OPFP-PBAA OPFP-RPDC Out N (PDC) Distribution Center (PDC) Unit 2 Recirc Power Distribution Center RPDC2 RPDC2 Unit 2 Recirc Power OPFP-PBAA OPFP-RPDC Out N (PDC) Distribution Center (PDC) Radwaste Building Tank Room, minus 3ft RW-1 RW-01A Radwaste Building OPFP-PBAA 0PFP-RW-la IN Y Radwaste Building CFD Area, 23ft RW-1 RW-01B Radwaste Building OPFP-PBAA OPFP-RW-lb IN Y Radwaste Building Processing, 35ft RW-1 RW-01C Radwaste Building OPFP-PBAA OPFP-RW-lc IN Y Radwaste Building Processing, 47ft RW-1 RW-01D Radwaste Building OPFP-PBAA OPFP-RW-ld IN Y Radwaste Building Roof, 44ft RW-1 RW-01E Radwaste Building OPFP-PBAA OPFP-RW-le IN Y Radwaste Building Elevator Machinery RW1 RW-01F Radwaste Building OPFP-PBAA OPFP-RW-lf IN Y Room, 70ft Service Water Building Pump Area, 20ft SWl-1 SW1-01A Service Water Building OPFP-PBAA 0PFP-SW-la Out N Service Water Building Basement, 4ft SWI-1 SW1-01B Service Water Building OPFP-PBAA OPFP-SW-lb Out N Service Water Building Sump, minus 13ft SWl-1 SW1-01C Service Water Building OPFP-PBAA 0PFP-SW-lb Out N East Yard Open Area YARD EY Yard OPFP-PBAA OPFP-EY IN N Transformer Yard YARD TY Yard OPFP-PBAA OPFP-TY Out N Chlorination Building CLB CLB Chlorination Building OPFP-PBAA OPFP-EY Out N Unit 1 Condensate Transfer Pump House CTPH1 CTPH1 Unit 1 Condensate OPFP-PBAA OPFP-EY IN Y Transfer Pump House Unit 2 Condensate Transfer Pump House CTPH2 CTPH2 Unit 2 Condensate OPFP-PBAA OPFP-EY IN Y Transfer Pump House Duct Bank under East Yard DUCTBANK DUCTBANK DUCTBANK OPFP-PBAA OPFP-EY Out N Unit 1 HPCI C02 Bottle Room HCB1 HCB1 Unit Room1 HPCI C02 Bottle OPFP-PBAA OPFP-TY IN Y Unit 2 HPCI C02 Bottle Room HCB2 HCB2 Unit 2 HPCI C02 Bottle OPFP-PBAA OPFP-TY IN Y Room HVAC Cooling Towers HCT HCT HVAC Cooling Towers OPFP-PBAA OPFP-TY Out N Old NDE Shack NDE NDE Old NDE Shack OPFP-PBAA OPFP-EY Out N I Page 38 I IBSEPLARRev2 BSEP LAR Rev 2 Page 38

CP&L 4.0 Compliance with NFPA 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) YIN Stack Filter House SFH SFH Stack Filter House OPFP-PBAA OPFP-EY IN Y Stack Monitoring House SMH SMH Stack Monitoring House OPFP-PBAA OPFP-EY Out N Sewage Treatment Plant STP STP Sewage Treatment Plant OPFP-PBAA OPFP-TY Out N Unit 1 Valve Pit VP1 VP1 Unit 1 Valve Pit OPFP-PBAA OPFP-EY Out N Unit 2 Valve Pit VP2 VP2 Unit 2 Valve Pit OPFP-PBAA 0PFP-EY Out N Circwater Yard YARD CW Yard OPFP-PBAA OPFP-EY Out N Radwaste Loading Dock YARD RWLD Yard OPFP-PBAA OPFP-RW-lb IN Y Northwest Yard YARD NWY Yard OPFP-PBAA OPFP-TY Out N Radwaste Building Elevator RW-1 RW-01G Radwaste Building OPFP-PBAA OPFP-RW-lf Out Y Reactor Building Southwest Core Spray, RB1-1 RB1-01A Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-1a IN Y minus 17ft General Areas Reactor Building Northwest Core Spray, RB1-1 RB1-01B Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lb IN Y minus 17ft General Areas Reactor Building Northeast RHR Room, RB1-1 RB1-01C Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lc IN Y minus 17ft General Areas Reactor Building Southeast RHR Room, RB1-1 RB1D-01 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-ld IN Y minus 17ft General Areas Reactor Building Northeast RHR Heat RB1-1 RB1-O1E Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-1e IN Y Exchanger, 20ft RBI-1_B1-01 General Areas 1F-B PPRIl IN Reactor Building Southeast RHR Heat RB1-1 RBl-01F Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lf IN Y Exchanger, 20ft General Areas Reactor Building East Central, 20ff RB1-1 RB1- Unit 1 Reactor Building 1PFPRB 1PFP-RB1-lg N IN Y 01G(EC) General Areas 1PFP-RB1-lg S Reactor Building North Central, 20ff RB1-1 R1B1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-1g N IN Y 01G(NC) General Areas Reactor Building Northeast Corner, 20ff RB1-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lg N IN Y 01G(NE) General Areas Reactor Building Northwest Corner, 20ff R11-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lg N IN Y RsG(NW) General Areas Reactor Building South Central, 20ft RB1-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lg S IN Y _____________________________________ 1G(SC) General Areas

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CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Procedure Pre Pre-Plan Input Input RCA? Y/N (Screened) Reactor Building Southeast Corner, 2ft R131-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lg S IN Y 01G(SE) General Areas Reactor Building Southwest Corner, 20ft RB1-1 R131- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lg S IN Y 01(W General Areas Reactor Building East Central, 5Oft R131-1 RB1-01 H(EC) Unit 1 Reactor General Areas Building 1PFP-RB 1FR 1PFP-RB1-1h 1PBIhE E IN Y N Reactor Building North Central, 50ft RB1-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RBI-1 h E IN Y 01 H(NC) General Areas 1PFP-RBI-1h W Reactor Building Northeast Corner, 50ft R11-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lh E IN Y 01 H(NE) General Areas Reactor Building Northwest Corner, 50ft RB1-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lh W IN Y t01H(NW) General Areas Reactor Building Southeast Corner, 5Oft R131-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lh E IN Y 01 H(SE) General Areas 1FR PPB-hE N Reactor Building Southwest Corner, 50ft RB1-1 RB1- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lh W IN Y 01HSW General Areas Reactor Building West Central, 50ft RB1-1 R131- Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lh W IN Y 01 H(WC) General Areas _____ Reactor Building RWCU Access Room, 77ft RB1-1 RB1-011 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-12 IN Y General Areas Building Reactor Building West, 80ft RB1-1 RB1-01J Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lj W IN Y General Areas Reactor Building East, 8Oft RB1-1 RBI-O1K Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lk IN Y General Areas Reactor Building Spent Fuel Pool, 117ft R131-1 RB1-O1L Unit 1 Reactor Building 1PFP-RB 1PFP-RBI-1m IN Y General Areas Reactor Building Refueling Floor, 117ft R131-1 RBl-01M Unit 1 Reactor Building 1PFP-RB 1PFP-RBI-1m IN Y General Areas Reactor Buildin HPCI Rof Mezzanine 5f RB1 R11N Unit 1 Reactor Building 1PFPRB 1PFP-RB1-le IN Y g 0 eae, 5- - General Areas 1PFP-RB1-lf Reactor Building ECCS Tunnel Roof, 36ft RB1-1 RB1-010 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lo IN Y _____________________ I_____ I_____ General Areas I__ I__________________ I__ I Page 40 I IBSEPLARRev2 BSEP LAR Rev 2 Page 40

CP&L 4.0 Compliance with NFPA 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre Input RCA? Procedure Pre-Plan Input Y/N (Screened) Reactor Building HPCI Room, minus 7ft R131-1 RBl-02 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-2 IN Y General Areas Reactor Building Drywell and Torus RB1-1 RB1-03 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-3 IN Y General Areas Reactor Building MSIV Pit, 50ft R131-1 RBl-04 Unit 1 Reactor Building 1PFP-RB3 1PFP-RB1-4 IN Y General Areas Reactor Building HP Field Office, RB1-1 RB1-05 Unit 1 Reactor Building 1PFP-RB 1PFP-RBI-lgS IN Y Decontamination Room, 20ft General Areas Reactor Building Drywell Entry, 20ft RBI-1 RB1-07 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-7 IN Y General Areas Reactor Building TIP Room, 20ft RB1-1 RB1 -08 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-8 IN Y General Areas Reactor Building Elevator Shaft R11-1 RBl-09 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-9 IN Y General Areas Reactor Building RWCU Pump and Heat R11 RBl10 Unit 1 Reactor Building 1PFP-RB lPFP-RBl-10 IN Y Exchanger Room, 50ft General Areas Reactor Building New Fuel Vault, 7ft R131-11 RBl-11 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-16 IN Y General Areas Reactor Building RWCU Backwash Tank R11 RBl12 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-12 IN Y Room, 77ft General Areas Reactor Building CRD Repair Room, 80ft RB1-1 RB1-13 Unit 1 Reactor Building IPFP-RB 1PFP-RB1-13 IN Y East General Areas Reactor Building Skimmer Surge Tank Vault, RB1-1 RB1-14 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-14 IN Y 117ft General Areas Reactor Building Elevator Machinery Room, RB1-1 RB1-15 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-15 IN Y 133ft General Areas Reactor Building 1A RWCU Filter Pit, 11 7ft RB1-1 RBl-16 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-16 IN Y General Areas Reactor Building 1B RWCU Filter Pit, 117ft RB1-1 RB1-17 Unit 1 Reactor Building 1PFP-RB lPFP-RBl-16 IN Y General Areas Reactor Building Supply Room, 98ft RB1-1 RBl-18 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lp IN Y General Areas Reactor Building Platform, 98ft RB1-1 RBl-18GA Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-lp IN Y General Areas 1PFP-RB1-19 I Page 41 I I BSEP LAR Rev 2 BSEPLARRev2 Page 41

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Procedure Fre Pre-Plan Input Input RCA? Y/N (Screened) Reactor Building Clothing Change room, 98ft R11-1 RBI-19 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-19 IN Y ReactorBuilingClothingChangeroom 98ftRBI-1 RBGeneral Areas Reactor Building RWCU Valve Room, 77ft RB1-1 RBl-20 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-12 IN Y General Areas Reactor Building Resin Storage Room, 80ft RBl21 Unit 1 Reactor Building 1PFP-RB 1PFP-RB1-21 IN Y East RBI1 General Areas Reactor Building ECCS Mini Steam Tunnel, RB1-6 RB1-06 Mini Steam Tunnel 1PFP-RB 1PFP-RB1-6 IN Y 20ft Electrical Tunnel, 9ft TB1 ET Turbine Building General 1PFP-TB 1PFP-TB1-ET IN Y Areas Pipe Tunnel, -3ft TB1 PT Turbine Building General 1PFP-TB OPFP-RW-1a IN Y Areas Unit 1 TB Breezeway South, 20ft TB1 TB1-01A Turbine Building General 1PFP-TB 1PFP-TB1-1 IN Y Areas Unit 1 TB Breezeway North, 20ft TB1 TBl-01B Turbine Building General 1PFP-TB 1PFP-TB1-1 IN Y Areas Unit 1 TB Mechanical Vacuum Pump Area, TB1 TB1-01C Turbine Building General 1PFP-TB 1PFP-TB1-1c IN Y 20ft Areas Unit 1 TB Air Compressor Area, 20ft TB1 TB1-1 D Turbine Building General 1PFP-TB 1PFP-TB1-ld IN Y Areas Unit 1 TB 2A Air Dryer Area, 20ft TB1 TB1-01E Turbine Building General 1PFP-TB 1PFP-TB1-1e IN Y Areas Unit 1 TB 4KV Switchgear Area, 20ft TB1 TB1-01F Turbine Building General 1PFP-TB 1PFP-TB1-lf IN Y Areas Unit 1 TB3 Hydrogen Seal Oil Area, 2Oft TB1 TB1-1G Turbine Building General 1PFP-TB 1PFP-TB1-lg IN Y Unit_1_TBHydrogenSealOilArea,_20ft TB1G Areas Unit 1 TB Condensate Pump Area, 20ft TB1 TB1-01H Turbine Building General 1PFP-TB 1PFP-TB1-lh IN Y Areas Unit 1 TB 1A Reactor Recirc MG Set Room, T11 T131-il Turbine Building General 1PFP-TB 1PFP-TB1-li & IN Y 38ft Areas Unit 1 TB 1B Reactor Recirc MG Set Room, TB1 TB1-01J Turbine Building General 1PFP-TB 1PFP-TB1-li &j IN Y 38ft Areas________ _ ____ Unit 1 TB South 38ft and 45ft TB1 TB1-01K Turbine Building General 1PFP-TB 1PFP-TB1-lk IN Y I___ ___ IAreas II I Page 42 I BSEPLARRev2 BSEP LAR Rev 2 Page 42

CP&L 4.0 Compliance with NFPA 806 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Relaase RCA? Procedure Pre-Plan Input Y/N (Screened) Unit 1 TB Supply Fan Room 55ff T11 TB1-01L Turbine Building General 1PFP-TB 1PFP-TBl-1l IN Y Areas Unit 1 TB Main Turbine Front Standard Area, TB1 TB1-01M Turbine Building General 1PFP-TB 1PFP-TB1-lm IN Y 70ft Areas Unit 1 TB Main Turbine and MSR Area, 70ft TB1 TB1-01N Turbine Building General 1PFP-TB 1PFP-TB1-1n IN Y Areas Unit 1 TB Main Generator and Exciter area, TB1 TB1-010 Turbine Building General 1PFP-TB 1PFP-TB1-lo IN Y 70ft Areas Unit 1 TB 1B SJAE Room, 20ft TB1 TB1-02 Turbine Building General 1PFP-TB 1PFP-TB1-2 IN Y Areas Unit 1 TB 1A SJAE Room, 20ff TB1 TBl-03 Turbine Building General 1PFP-TB 1PFP-TB1-3 IN Y UtT1SARm2tB- Areas Unit 1 TB 1B RFPT Room, 20ff T11 TB1-04 Turbine Building General 1PFP-TB 1PFP-TB1-4 IN Y UtT1 F Rm2T11 Areas Unit 1 TB 1A RFPT Room, 20ft TB1 TBl-05 Turbine Building General 1PFP-TB 1PFP-TB1-5 IN Y Areas Unit 1 TB Condensate Booster Pump Room, TB1 TB1-06 Turbine Building General 1PFP-TB 1PFP-TB1-6 IN Y 20ft Areas Unit 1 TB Heater Drain Pump Room, 9ff TB1 TBl-07 Turbine Building General 1PFP-TB 1PFP-TB1-7 IN Y U1B a Diu R ,t11 Areas Unit 1 TB Condenser Bay Area, 20ft TB1 TB1-08A Turbine Building General 1PFP-TB 1PFP-TB1-8a IN Y Areas Unit 1 TB Condenser Pit East Area, 20ft TB1 TB1-08B Turbine Building General 1PFP-TB 1PFP-TB1-8b IN Y Areas Unit 1 TB Condenser Pit West Area, 20ff T11 TBl-08C Turbine Building General 1PFP-TB 1PFP-TB1-8c IN Y Areas Unit 1 TB Condenser Bay Area, 45ff TB1 TBl-08D Turbine Building General 1PFP-TB 1PFP-TB1-8d IN Y Areas Unit 1 TB Condenser Pit East Area, 45ff T11 TBl-08E Turbine UTCd r Et a f B1Areas Building General 1PFP-TB 1PFP-TB1-8e IN Y Unit 1 TB Condenser Pit West Area, 45ft TB1 TB1-08F Turbine Building General 1PFP-TB 1PFP-TB1-8f IN Y Areas I Unit 1 TB EHC and Lube Oil Room, 20ff TB1 TB1-09A Turbine Building General 1PFP-TB 1PFP-TB1-9a IN Y Unit__1_TBEHCandLubeOilRoom,_20ftTB1_TB1-09A Areas I Page 43 I IBSEPLARRev2 BSEP LAR Rev 2 Page 43

CP&L 4.0 Compliance with NFPA. 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre Input RCA? Procedure Pre-Plan Input YIN (Screened) Unit 1 TB EHC and Lube Oil Room, 45ff T11 TBl-09B Turbine Building General 1PFP-TB 1PFP-TB1-9b IN Y Areas Unit 1 TB Exhaust Fan Room, 45ff T11 TBl-10 Turbine Building General 1PFP-TB 1PFP-TBl-10 IN Y Areas Unit 1 TB A Train HP Feedwater heater T11 TBl-12 Turbine Building General 1PFP-TB 1PFP-TB1-12 IN Y Room, 45ft Areas Unit 1 TB B Train HP Feedwater heater T11 TBl13 Turbine Building General 1PFP-TB 1PFP-TB1-13 IN Y Room, 45ft Areas Unit 1 TB Receiving Area T11 TBl-15 Turbine Building General 1PFP-TB 0PFP-013 IN Y Areas Unit 1 Heater Bay Roof YARD TB11- Yard 1PFP-TB 1PFP-TBI-lm IN Y HBROOF Yad1F-B PFTB-m IY Turbine Building 1 Dragon's Breath TBI TB1-DB Turbine Building General 1PFP-TB lPFP-TB1-12 IN Y Areas Reactor Building Southwest Core Spray, RB2-1 RB2-01A Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1a IN Y minus 17ft General Areas Reactor Building Northwest Core Spray, RB2-1 RB2-01B Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1b IN Y minus 17ft General Areas Reactor Building Northeast RHR Room, RB2-1 RB2-01C Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1c IN Y minus 17ft General Areas Reactor Building Southeast RHR Room, RB2-1 RB2-01D Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-ld IN Y minus 17ft General Areas Reactor Building Northeast RHR Heat RB2-1 R12-01E Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-Ie IN Y Exchanger, 20ft General Areas Reactor Building Southeast RHR Heat R12-1 R12-01F Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lf IN Y Exchanger, 20ft General Areas Reactor Building East Central, 20ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg N IN Y RcBln aC rfB 01G(EC) General Areas 2PFP-RB2-lg S Reactor Building North Central, 20ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg N IN Y Rcrud NrCnlB 01G(NC) General Areas Reactor Building Northeast Corner, 20ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg N IN Y 01G(NE) General Areas Reactor Building Northwest Corner, 20ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg N IN Y Reactor__BuildingNorthwestCorner,_20ft _RB2-1_ 01G(NW) General Areas I Page 44 I I BSEP LAR Rev 2 BSEPLARRev2 Page 44

CP&L 4.0 Compliance with NFPA 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) Y/N Reactor Building South Central, 20ft RB2-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg S IN Y 01G(SC) General Areas Reactor Building Southeast Corner, 2ft RB2-1 R1B2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg S IN Y ReactorBuildingSoutheastCorner,_20ft RB-1 01G(SE) General Areas Reactor Building Southwest Corner, 20f RB2-1 R1B2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg S IN Y ReactorBuildingSouthwestCorner,_20f RB2-1 01G(SW) General Areas Reactor Building East Central, 50ff RB2-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lh E IN Y Reactor Building East Central, 50ft RB2-1 01H(EC) General Areas Reactor Building North Central, 50ft RB2-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1 h E IN Y 01H(NC) General Areas 2PFP-RB2-1 h W Reactor Building Northeast Corner, 50ft RB2-1 R12- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lh E IN Y 01H(NE) General Areas Reactor Building Northwest Corner, 50ft RB2-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1h W IN Y 01H(NW) General Areas Reactor Building Southeast Corner, 50ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1h E IN Y 01 H(SE) General Areas Reactor Building Southwest Corner, 50ft RB2-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1h W IN Y 01H(SVM) General Areas Reactor Building West Central, 50ff R12-1 RB2- Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1 h W IN Y 01 H(WC) General Areas Reactor Building RWCU Access Room, 77ff R12-1 R12-011 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-12 IN Y General Areas Reactor Building West, 80ff R12-1 R12-01J Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lj W IN Y General Areas Reactor Building East, 80ff 132-1 RB2-01 K Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1k IN Y General Areas Reactor Building Spent Fuel Pool, 117ff R132-1 RB2-OL Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lm IN Y General Areas Reactor Building Refueling Floor, 117ff RB2-1 R32-01 M Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1 m IN Y i General Areas Reactor Building HPCI Roof Mezzanine, 5ff 32-1 R132-01N Unit 2 Reactor Building 2PFPRB 2PFP-RB2-le IN Y General Areas 2PFP-RB2-lf I Page 45 I I BSEP LAR Rev 2 BSEPLARRev2 Page 45

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In FieFire Release RA Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Procedure Pre Pre-Plan Input Input RCA? Y/N (Screened) Reactor Building ECCS Tunnel Roof, 36ft RB2-1 RB2-010 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1o IN Y General Areas Reactor Building HPCI Room, minus 17ft RB2-1 RB2-02 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-2 IN Y ReactorBuidingHPlRoom,_minus_17f RB2-1 _ RB2-02 _ General Areas Reactor Building Drywell and Torus R12-1 RB2-03 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-3 IN Y Reactor__BuildingDrywellandTorusB2-1 RB2-03 _ General Areas Reactor Building MSIV Pit, 50ft RB2-1 RB2-04 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-4 IN Y General Areas Reactor Building HP Field Office, R12-1 RB2-05 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lg IN Y Decontamination Room, 20ft RB2-1 RB2-05 General Areas Reactor Building Drywell Entry, 20ft R12-1 RB2-07 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-7 IN Y General Areas Reactor Building TIP Room, 20ft RB2-1 RB2-08 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-8 IN Y General Areas Reactor Building Elevator Shaft R12-1 RB2-09 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-9 IN Y RaoBlnEetS2 RGeneral Areas Reactor Building RWCU Pump and Heat RB2-1 R12-1 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1O IN Y Exchanger Room, 50ft General Areas Reactor Building New Fuel Vault, 117ft RB2-1 RB2-11 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1m IN Y General Areas Reactor Building RWCU Backwash Tank RB2-1 RB2-12 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-12 IN Y Room, 77ft RB2-1 RB2-12 General Areas Reactor Building CRD Repair Room, 80ft RB2-1 RB2-13 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-13 IN Y East General Areas Reactor Building Skimmer Surge Tank Vault, RB2-1 RB2-14 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-14 IN Y 117ft General Areas Reactor Building Elevator Machinery Room, RB2-1 RB2-15 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-15 IN Y 133ft General Areas Reactor Building 2A RWCU Filter Pit, 17ft R132-1 RB2-16 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-16 IN Y General Areas Reactor Building 2B RWCU Filter Pit, 117ft RB2-1 RB2-17 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-16 IN Y General Areas Reactor Building Supply Room, 98ft R12-1 R12-18 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-lp IN Y Reactor__Building __Supply __Room,__98ft__RB2-1 _ R-General Areas I Page 46 I I BSEP LAR Rev 2 BSEPLARRev2 Page 46

CP&L 4.0 Compliance with NFPA 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) Y/N Reactor Building Platform, 98ft RB2-1 RB2-18GA Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-1 p IN Y _________________________________General Areas____________ Reactor Building Clothing Change room, 98ff RB2-1 R12-19 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-19 IN Y ReactorBuildingClothingChangeroom, 9fR2RB 9 General Areas Reactor Building Resin Storage Room, 80ft R12-20 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-20 IN Y East RB21 General Areas Reactor Building RWCU Valve Room, 77ff RB2-1 R12-21 Unit 2 Reactor Building 2PFP-RB 2PFP-RB2-12 IN Y ReactorBuildingRWCUValveRoom,_77ft RB2-1_RB2-21 General Areas Reactor Building ECCS Mini Steam Tunnel, RB2-6 RB2-06 Mini Steam Tunnel 2PFP-RB 2PFP-RB2-6 IN Y 20ft Unit 2 TB Breezeway North TB1 TB2-O1A Turbine Building General 2PFP-TB 2PFP-TB2-1 IN Y Unit_2_TBBreezewayNorth_ TB1 TB2-01A Areas Unit 2 TB Breezeway South TB1 T12-01B Turbine Building General 2PFP-TB 2PFP-TB2-1 IN Y Unit__2_TB __BreezewaySouth_ TB1 TB2-01B Areas Unit 2 TB Mechanical Vacuum Pump Area, TB1 TB2-01C Turbine Building General 2PFP-TB 2PFP-TB2-1c IN Y 20ft Areas Unit 2 TB Air Compressor Area, 20ff TB1 T32-01D Turbine Building General 2PFP-TB 2PFP-TB2-ld IN Y ________________________________Areas __________ Unit 2 TB 2A Air Dryer Area, 20ft TB1 TB2-01E Turbine Building General 2PFP-TB 2PFP-TB2-1e IN Y Areas Unit 2 TB 4KV Switchgear Area, 20ff T131 TB2-1F Turbine Building General 2PFP-TB 2PFP-TB2-lf IN Y Unit_2_TB_4KVSwitchgearArea,_20ft _ TB1 T20F Areas Unit 2 TB Hydrogen Seal Oil Area, 20ff TB1 TB2-1 G Turbine Building General 2PFP-TB 2PFP-TB2-lg IN Y Areas Unit 2 TB Condensate Pump Area, 20ft TB1 TB2-01 H Turbine Building General 2PFP-TB 2PFP-TB2-1h IN Y Areas Unit 2 TB 2B Reactor Recirc MG Set Room, TB1 TB2-011 ATurbine Building General 2PFP-TB 2PFP-TB2-1i &j IN Y 38ff _____ Areas________ Unit 2 TB 2A Reactor Recirc MG Set Room, TB1 TB2-01J Turbine Building General 2PFP-TB 2PFP-TB2-1i &j IN Y 38ff _____ Areas________ Unit 2 TB North 38ft and 45ft TB1 TB2-01K Turbine Areas Building General 2PFP-TB 2PFP-TB2-1k I IN Y Unit 2 TB Supply Fan Room 55ff TB1 TB2-O L Turbine Building General 2PFP-TB 2PFP-TB2-11 IN Y Unit_2_TBSupplyFanRoom_55ft _ TB1 TB2-01LAreas I Page 47 I I BSEPLARRev2 I BSEP LAR Rev 2 Page 47

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Rad In Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Fire Release RCA? Procedure Pre-Plan Input Y/N (Screened) Unit 2 TB Main Turbine Front Standard Area, TB1 TB2-01M Turbine Building General 2PFP-TB 2PFP-TB2-lm IN Y 70ft Areas Unit 2 TB Main Turbine and MSR Area, 70ft TB1 TB2-01N Turbine Building General 2PFP-TB 2PFP-TB2-1n IN Y Areas Unit 2 TB Main Generator and Exciter area, TB1 TB2-010 Turbine Building General 2PFP-TB 2PFP-TB2-lo IN Y 70ft Areas Unit 2 TB 2B SJAE Room, 20ft TB1 TB2-02 Turbine Building General 2PFP-TB 2PFP-TB2-2 IN Y Areas Unit 2 TB 2A SJAE Room, 20ft TB1 TB2-03 Turbine Building General 2PFP-TB 2PFP-TB2-3 IN Y Areas Unit 2 TB 2B RFP Room, 20ft TB1 TB2-04 Turbine Building General 2PFP-TB 2PFP-TB2-4 IN Y Areas 2PFP-TB 2PFP-TB2-4_ IN_ Unit 2 TB 2A REP Room, 20ft TB1 T12-05 Turbine Building General 2PFP-TB 2PFP-TB2-5 IN Y Unit__2_TB_2ARFPRoom,_20ft _ TB1 TB2-05_Areas Unit 2 TB Condensate Booster Pump Room, TB1 T1206 Turbine Building General 2PFP-TB 2PFP-TB2-6 IN Y 20ft ______Areas Unit 2 TB Heater Drain Pump Room, 20ft TB1 TB2-07 Turbine Areas Building General 2PFP-TB 2PFP-TB2-7 IN Y Unit 2 TB Condenser Bay Area, 20ft TB1 TB2-08A Turbine Building General 2PFP-TB 2PFP-TB2-8a IN Y ______________________________________Areas Unit 2 TB Condenser Pit East Area, 20ft TB1 TB2-08B Turbine Areas Building General 2PFP-TB 2PFP-TB2-8b IN Y Unit 2 TB Condenser Pit West Area, 20ft TB1 TB2-08C Turbine Building General 2PFP-TB 2PFP-TB2-8c IN Y Unit_2_TBCondenserPitWestArea,_20ft TB1_TB2-08C Areas Unit 2 TB Condenser Bay Area, 45ft TB1 TB2-08D Turbine Building General 2PFP-TB 2PFP-TB2-8d IN Y Areas Unit 2 TB Condenser Pit East Area, 45ft T11 TB2-08E Turbine Building General 2PFP-TB 2PFP-TB2-8e IN Y Areas Unit 2 TB Condenser Pit West Area, 45ft TB1 TB2-08F Turbine Building General 2PFP-TB 2PFP-TB2-8f IN Y Areas Unit 2 TB EHC and Lube Oil Room, 20ft TB1 TB2-09A Turbine Building General 2PFP-TB 2PFP-TB2-9a IN Y Areas Unit 2 TB EHC and Lube Oil Room, 45ft TB2-09B Turbine Building General 2PFP-TB 2PFP-TB2-9b IN Y Unit_2_TBEHCandLubeOilRoom,_45ftTB1 TB2-09B Areas I I___ I Page 48 I I BSEP BSEPLARRev2 LAR Rev 2 Page 48

CP&L 4.0 Compliance with NFPA 805 Requirements Fire FieFire Rad Release RAIn Fire Zone Description Fire Area Fire Zone Fire Area Description Pre-Plan Pre-Plan Input RCA? Procedure Pr-lnIpt (Screened) Y/N Unit 2 TB3 Exhaust Fan Room, 45ft TB1 T32-1 Turbine Building General 2PFP-TB 2PFP-TB2-10 IN Y Areas Unit 2 TB A Train HP Feedwater heater TB1 TB212 Turbine Building General 2PFP-TB 2PFP-TB2-12 IN Y Room, 45ft Areas 2FTPPB-2 N Unit 2 TB B Train HP Feedwater heater TB1 T1213 Turbine Building General 2PFP-TB 2PFP-TB2-13 IN Y Room, 45ft TB1_TB2-13 Areas Turbine Building Elevator Shaft TB1 T12-14 Turbine Building General 2PFP-TB 2PFP-TB2-14 IN Y TbeuiEeoSfBT- Areas Turbine Building Elevator Machinery Room TB1 TB2-15 Turbine Building General 2PFP-TB 2PFP-TB2-15 IN Y Areas Turbine Building Laydown Area TB1 TB2-16 Turbine Building General 2PFP-TB 2PFP-TB2-16 IN Y _________ Areas______ Unit 2 Heater Bay Roof YARD TB2-HBROOF Yard 2PFP-TB 2PFP-TB2-1m IN Y Turbine Building 2 Dragon's Breath T11 TB2-DB Turbine Building General 2PFP-TB 2PFP-TB2-12 IN Y T iBd 2rnBtBB Areas Unit 1 Control Building Roof YARD CB-ROOF1 Yard OPFP-013 *New PFP IN Y Unit 2 Control Building Roof YARD CB-ROOF2 Yard OPFP-013 *New PFP IN Y Reactor Building 1 Roof YARD RB1-ROOF Yard OPFP-013 *New PFP IN Y I Page 49 I IBSEPLARRev2 BSEP LAR Rev 2 Page 49

CP&L 4.0 Compliance with NFPA 805 Requirements Fire Brigade Training Plan Review; BSEP has completed transition of its fire brigade and site incident commander lesson plans to a fleet standard, NFPA 600 compliant format, aligning with NFPA 805, Section 3.4.1. Attributes are included within the new NFPA 600 lesson plans to address the Radioactive Release objectives. Lesson plan topics are technical skill-set based rather than fire area specific. As such, discussion points were noted for the topics applicable to, or having potential impact to radioactive release due to firefighting activities. Discussion points are included regarding containment and monitoring of potentially contaminated fire suppression agents and products of combustion for the following lesson plan topical areas;

      "   Safety and Orientation
      "   Personnel Protective Equipment
      "   Fire Hose
      "   Forcible Entry
      "   Ventilation
      "   Overhaul
      "   Fire Attack Engineering Controls Review; The review panel determined Engineering Controls are adequate to ensure that radioactive materials (i.e., radiation) generated as a direct result of fire suppression activities is contained and monitored prior to release to unrestricted areas such that such release would be as low as reasonably achievable and would not exceed applicable 10 CFR, Part 20 limits. Engineering controls such as use of forced air ventilation and damming for fire suppression agent run-off was considered during review of fire pre-plans, for areas in which this is the anticipated response identified in the pre-fire plan. No new engineering controls were identified or established as a result of this review, and all present controls are as currently in place under the approved pre-transitional fire protection program.

Documentation; Results of the radioactive release reviews described above have been documented in summary format in Attachment E. Open Items identified in the review process will be incorporated into the indicated fire pre-plans. 4.5 Fire PRA and Performance-Based Approaches RI-PB evaluations are an integral element of an NFPA 805 fire protection program. Key parts of RI-PB evaluations include:

      " A Fire PRA (i.e., discussed in Section 4.5.1 and Attachments U, V, and W).
      " NFPA 805 Performance-Based Approaches (i.e., discussed in Section 4.5.2).

l Page 50 I IBSEP LAR Rev 2 BSEPLARRev2 Page 50

CP&L 4.0 Compliance with NFPA 805 Requirements 4.5.1 Fire PRA Development and Assessment In accordance with the guidance in RG 1.205, a Fire PRA model was developed for BSEP in compliance with the requirements of Part 4, "Requirements for Fires At Power PRA," of the ASME and ANS combined PRA Standard, ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application," (i.e., hereafter referred to as Fire PRA Standard). CP&L conducted a peer review by independent industry analysts in accordance with RG 1.200 prior to a risk-informed submittal. The resulting fire risk assessment model is used as the analytical tool to perform Fire Risk Evaluations during the transition process. Section 4.5.1.1 describes the Internal Events PRA model. Section 4.5.1.2 describes the Fire PRA model. Section 4.5.1.3 describes the results and resolution of the peer review of the Fire PRA, and Section 4.5.1.4 describes insights gained from the Fire PRA. 4.5.1.1 Internal Events PRA The Brunswick Unit 1 and 2 base internal events PRA (i.e., Calculation BNP-PSA-030) was the starting point for the Fire PRA. Attachment U provides a discussion of the internal events PRA and the results and disposition of the most recent peer review. 4.5.1.2 Fire PRA The internal events PRA was modified to capture the effects of fire both as an initiator of an event and as a potential failure mode of affected circuits and individual targets. The Fire PRA was developed using the guidance for Fire PRA development in NUREG/CR-6850/EPRI TR 1011989, approved FAQs, and EPRI TR 1016735. The Fire PRA quality and results are discussed in the subsequent sections and in Attachments V and W, respectively. Fire Model Utilization in the Application RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance on documenting the fire models used, and justifying that these fire models and methods are acceptable for use in performance-based analyses when performed by qualified users, have been verified and validated, and are used within their limitations and with the rigor required by the nature and scope of the analyses. As part of the NFPA 805 transition, fire modeling was performed as part of the Fire PRA development (i.e., NFPA 805 Section 4.2.4.2) and, therefore, maximum expected fire scenario (MEFS)/limiting fire scenario (LFS) were not analyzed separately. RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis. The following fire models were used:

      " Fire Dynamics Tools (FDT's)
      " Consolidated Model of Fire and Smoke Transport (CFAST)
      " Fire Dynamics Simulator (FDS) l Page 51 I IBSEP  LAR Rev 2 BSEPLARRev2                                                                         Page 51

CP&L 4.0 Compliance with NFPA 805 Requirements The approach taken at BSEP to simplify the analysis process incorporates features of several fire model tools covered by NUREG-1824, as well as additional features. The approach is collectively referred to as the Fire Modeling Generic Treatments. The analysis basis and Verification and Validation (V&V) documentation was provided in a proprietary Hughes Associates, Inc. report to the NRC on January 24, 2008. The report entitled "Generic Fire Modeling Treatments" is effectively a technical reference guide, a user's guide, and the V&V basis. The use of the Generic Treatments in specific applications at BSEP falls within their limitations as described in the "Generic Fire Modeling Treatments". In addition to the generic fire modeling treatments that were used in the hazard analysis, several calculations were produced that used FDS, CFAST, and the FDT's as documented in NUREG-1824. The acceptability of the use of these fire models is included in Attachment J. 4.5.1.3 Results of Fire PRA Peer Review The Brunswick Unit 1 and 2 Fire PRA (i.e., Calculations BNP-PSA-080 and BNP-PSA-082) was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4 and Regulatory Guide 1.200, revision 2. The results (i.e., Supporting Requirement capability assessments and Facts & Observations (F&Os)) documented in the February 2012 Fire PRA peer review report, and subsequent focused scope peer review reports, were used to support the Fire PRA for the NFPA 805 application. The Fire PRA update addressed the Supporting Requirement assessed deficiencies (i.e., Not Met or Capability Category I (CC I)). Completion of recommendations related to Supporting Requirement assessments and 'Finding' F&Os results in a Capability Category II assessment for the associated Supporting Requirements. Some items are not completed at this time and are deferred. These items have been dispositioned for the potential impact on the Fire PRA and the application. The results of the peer review are summarized in Attachment V. 4.5.1.4 Risk Insights Risk insights were documented as part of the development of the Fire PRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for fire PRA development and is useful in identifying the areas of the plant where fire risk is greatest. A review of the fire initiating events that collectively represent 95% of the calculated fire risk is included as Attachment W. 4.5.2 Performance-Based Approaches NFPA 805 outlines the approaches for performing performance-based analyses. As specified in Section 4.2.4, there are generally two types of analyses performed for the performance-based approach:

       " Fire Modeling (i.e., NFPA 805, Section 4.2.4.1).
       " Fire Risk Evaluation (i.e., NFPA 805, Section 4.2.4.2).

I Page 52 I IBSEP BSEPLARRev2 LAR Rev 2 Page 52

CP&L 4.0 Compliance with NFPA 805 Requirements 4.5.2.1 Fire Modeling Approach The fire modeling approach was not utilized for demonstrating compliance with NFPA 805 for BSEP. 4.5.2.2 Fire Risk Approach Overview of Evaluation Process The Fire Risk Evaluations were completed as part of the BSEP NFPA 805 transition. These Fire Risk Evaluations were developed using the process described below. This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These are summarized in Table 4-1. Table 4-1 Fire Risk Evaluation Guidance Summary Table Document Section(s) Topic NFPA 805 2.2(h), 4.2.4, A.2.2(h), A.2.4.4, D.5 Change Evaluation (2.2(h), 2.2.9, 2.4.4 A.2.2(h), A.2.4.4, D.5) Risk of Recovery Actions (4.2.4) Use of Fire Risk Evaluation (4.2.4.2) NEI 04-02 Revision 2 4.4, 5.3, Appendix B, Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (App. I), No specific discussion of Fire Risk Evaluation RG 1.205 Revision 1 C.2.2.4, C.2.4, C.3.2 Risk Evaluations (C.2.2.4) Recovery Actions (C.2.4) During the transition to NFPA 805, variances from the deterministic approach in Section 4.2.3 of NFPA 805 were evaluated using a Fire Risk Evaluation per Section 4.2.4.2 of NFPA 805. A Fire Safety Analysis was performed for each fire area. For areas containing variances from the deterministic requirements (VFDRs) of Section 4.2.3 of NFPA 805, a Fire Risk Evaluation was performed for each fire area containing VFDRs. Ifthe Fire Risk Evaluation meets the acceptance criteria, this is confirmation that a success path effectively remains free of fire damage and that the performance-based approach is acceptable per Section 4.2.4.2 of NFPA 805. The Fire Risk Evaluation process consists of the following steps (Figure 4-7 depicts the Fire Risk Evaluation process used during transition. This is generally based on FAQ 07-0054 Revision 1: Step 1 - Preparation for the Fire Risk Evaluation.

     " Definition of the Variances from the Deterministic Requirements. The definition of the VFDR includes a description of problem statement and the section of NFPA 805 that is not met, type of VFDR (e.g., separation issue or degraded fire protection system), and proposed evaluation per applicable NFPA 805 section.
     " Preparatory Evaluation - Fire Risk Evaluation Team Review. Using the information obtained during the development of the NEI 04-02 B-3 Table and the Fire PRA, a team review of the VFDR was performed. Depending on the scope and complexity of the VFDR, the team may include the Safe shutdown/NSCA I

I BSEP LAR Rev 2 Page 53

CP&L 4.0 Compliance with NFPA 805 Requirements Engineer, the Fire Protection Engineer, and the Fire PRA Engineer. The purpose and objective of this team review was to address the following; o Review of the Fire PRA modeling treatment of VFDR o Ensure discrepancies were captured and resolved Step 2 - Performed the Fire Risk Evaluation The Evaluator coordinated as necessary with the Safe shutdown/NSCA Engineer, Fire Protection Engineer and Fire PRA Engineer to assess the VFDR using the Fire Risk Evaluation process to perform the following: o Change in Risk Calculation with consideration for additional risk of recovery actions and required fire protection systems and features due to fire risk. o Fire area change in risk summary Step 3 - Reviewed the Acceptance Criteria N The acceptance criteria for the Fire Risk Evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are ACDF and ALERF. The qualitative factors are defense-in-depth and safety margin. o Risk Acceptance Criteria. The transition risk evaluation was measured quantitatively for acceptability using the ACDF and ALERF criteria from RG 1.174, as clarified in RG 1.205, Regulatory Position 2.2.4. o Defense-in-Depth. A review of the impact of the change on defense-in-depth was performed, using the guidance NEI 04-02. NFPA 805 defines defense-in-depth as:

            -   Preventing fires from starting
            -   Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage
            -   Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the defense-in-depth requirement was considered to be satisfied if the proposed change does not result in a substantial imbalance among these elements (or echelons). The review of defense-in-depth was qualitative and addressed each of the elements with respect to the proposed change. Defense-in-depth was performed on a fire area basis. Fire protection features and systems relied upon to ensure defense-in-depth were identified as a result of the assessment of defense-in-depth. o Safety Margin Assessment. A review of the impact of the change on safety margin was performed. An acceptable set of guidelines for making that assessment is summarized below. Other equivalent acceptance guidelines may also be used. ! I BSEP LAR Rev 2 Page 54

CP&L 4.0 Compliance with NFPA 805 Requirements

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

The requirements related to safety margins for the change analysis are described for each of the specific analysis types used in support of the Fire Risk Evaluation (FRE). I Page 55 I IBSEP BSEPLARRev2 LAR Rev 2 Page 655

CP&L 4.0 Compliance with NFPA 805 Requirements Prepare for Fire Risk Evaluation Discuss and Document in Determine How to Model Fire PRA and Fire Risk the VFDR in the Fire PRA Evaluation Documentation Perform Fire Risk Evaluation Review of Acceptance Criteria Figure 4 Fire Risk Evaluation Process (NFPA 805 Transition) [Based on FAQ 07-0054 Revision 1] ! Page 56 IBSEPLARRev2 I BSEP LAR Rev 2 Page 56

CP&L 4.0 Compliance with NFPA 805 Requirements Results of Evaluation Process Disposition of VFDRs The BSEP existing post-fire SSA I NSCA and the NFPA 805 transition project activities have identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. These variances were dispositioned using the fire risk evaluation process. Each variance dispositioned using a Fire Risk Evaluation was assessed against the Fire Risk Evaluation acceptance criteria of ACDF and ALERF; and maintenance of defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in Attachment C. Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c). Risk Change Due to NFPA 805 Transition In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using Fire Risk Evaluations and the overall plant should be provided. Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area. RG 1.205 Section C.2.2.4.2 states in part "The total increaseor decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increasesresulting from previously approved recovery actions). The total risk increaseshould be consistent with the acceptanceguidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associatedwith previously approved recovery actions is greaterthan the acceptanceguidelines in Regulatory Guide 1.174, then the net change in total plant risk incurredby any proposed alternativesto the deterministic criteria in NFPA 805, Chapter4 (other than the previously approved recovery actions), should be risk neutralor representa risk decrease." The risk increases and decreases are provided in Attachment W. 4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states:

      "A monitoring program shall be established to ensure that the availabilityand reliabilityof the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria.

Monitoring shall ensure that the assumptions in the engineering analysis remain valid." m Page 57 I IBSEP LAR Rev 2 BSEPLARRev2 Page 57

CP&L 4.0 Compliance with NFPA 805 Requirements As part of the transition review, the adequacy of the inspection and testing program to address fire protection systems and equipment within plant inspection and the compensatory measures programs should be reviewed. In addition, the adequacy of the plant corrective action program in determining the causes of equipment and programmatic failures and minimizing their recurrence should also be reviewed as part of the transition to a risk-informed, performance-based licensing basis. 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program This section describes the process that will be utilized to implement the post-transition NFPA 805 monitoring program. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. See item for implementation in Attachment S. The monitoring process is comprised of four phases.

     " Phase 1 - Scoping
     " Phase 2 - Screening Using Risk Criteria
     " Phase 3 - Risk Target Value Determination
     " Phase 4 - Monitoring Implementation Figure 4-8 provides detail on the Phase 1 and 2 processes.

The results of these phases will be documented in the NFPA 805 Monitoring Program scoping document developed during implementation. Phase 1 - Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program:

     " Structures, Systems, and Components required to comply with NFPA 805, specifically:

o Fire protection systems and features

             - Required by the Nuclear Safety Capability Assessment
             - Modeled in the Fire PRA
             - Required by Chapter 3 of NFPA 805 o Nuclear Safety Capability Assessment equipment 4
             - Nuclear safety equipment
             - Fire PRA equipment
             - NPO equipment o Structures, systems and components relied upon to meet radioactive release criteria
     " Fire Protection Programmatic Elements 4 For the purposes of the NFPA 805 Monitoring, "NSCA equipment" is intended to include Nuclear Safety Equipment, Fire PRA equipment, and NPO equipment.

! I BSEP LAR Rev 2 Page 58

CP&L 4.0 Compliance with NFPA 805 Requirements Phase 2 - Screening Using Risk Criteria The equipment from Phase 1 scoping will be screened to determine the appropriate level of NFPA 805 monitoring. As a minimum, the SSCs identified in Phase 1 will be part of an inspection and test program and system/program health reporting. If not in the current program, the SSCs will be added in order to assure that the criteria can be met reliably. The following screening process will be used to determine those SSCs that may require additional monitoring beyond normal inspection and test program and system/program health reporting and will be documented in the NFPA 805 Monitoring Program scoping document.

1. Fire Protection Systems and Features Those fire protection systems and features identified in Phase 1 are candidates for additional monitoring in the NFPA 805 program commensurate with risk significance.

Risk significance is determined at the component, programmatic element, and/or functional level on an individual fire area basis. Compartments smaller than fire areas may be used provided the compartments are independent (i.e., share no fire protection SSCs). If compartments smaller than fire areas are used, the basis will be documented in the calculation, BNP-PSA-082. The Fire PRA is used to establish the risk significance based on the following screening criteria: Risk Achievement Worth (RAW) of the monitored parameter ->2.0 (AND) either Core Damage Frequency (CDF) x (RAW) _>1.OE-7 per year (OR) Large Early Release Frequency (LERF) x (RAW) ->1.OE-8 per year CDF, LERF, and RAW(monitored parameter) are calculated for each fire area. The 'monitored parameter' will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration) and will be documented in the calculation, BNP-PSA-082. Fire protection systems and features that meet or exceed the criteria identified above are considered High Safety Significant (HSS) and will be included in the NFPA 805 Monitoring Program The HSS fire protection systems and features not already monitored via an existing inspection and test program and/or in the existing system / program health reporting, as described in procedure EGR-NGGC-0010, will be added to the NFPA 805 Monitoring Program and documented in the NFPA 805 Monitoring Program scoping document.

2. Nuclear Safety Capability Assessment Equipment Required NSCA equipment, except the NPO scope, identified in Phase 1 will be screened for safety significance using the Fire PRA and the Maintenance Rule BSEP LAR Rev 2 Page 59

CP&L 4.0 Compliance with NFPA 805 Requirements guidelines differentiating HSS equipment from Low Safety Significant (LSS) equipment. The screening will also ensure that the Maintenance Rule functions are consistent with the required functions of the NSCA equipment. HSS NSCA equipment not currently monitored in Maintenance Rule will be added into Maintenance Rule. All NSCA equipment that are not HSS are considered LSS and need not be included in the monitoring program. For non-power operational modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement. Therefore, fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire prevention programs. Additional monitoring beyond inspection and test programs and system/program health reporting is not considered necessary.

3. SSCs Relied upon for Radioactive Release Criteria The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSCs relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (i.e., which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently low risk. Therefore, additional monitoring beyond inspection and test programs and system/program health reporting is not considered necessary.

4. Fire Protection Programmatic Elements Monitoring of programmatic elements is required in order to "assess the performance of the fire protection program in meeting the performance criteria". These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements. Programmatic aspects include:

      " Prompt Detection, including incipient detection fire watch and hot work fire watch
      " Transient Combustible Controls Program Violations against FIR-NGGC-0009
      " Fire Brigade Effectiveness including Fire Brigade Response Time, Fire Brigade Fire Drill, and Fire Brigade Fire Drill Objectives Monitoring of programmatic elements is more qualitative in nature since the programs do not lend themselves to the numerical methods of reliability and availability.

Therefore, monitoring is conducted using the existing program health programs. Fire protection health reports, self-assessments, regulator and insurance company reports provide inputs to the monitoring program. ! Page 60 I IBSEP LAR Rev 2 BSEPLARRev2 Page 60

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements Phase 3 - Risk Target Value Determination Failure criteria is established by an expert panel based on the required fire protection and nuclear safety capability SSCs and programmatic elements assumed level of performance in the supporting analyses established in Phase 2. Action levels are established for the SSCs at the component level, program level, or functionally through the use of the pseudo system or 'performance monitoring group' concept. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (i.e., 3 operating cycles). Since the HSS NSCA equipment have been identified using the Maintenance Rule guidelines, the associated equipment specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions. When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions. Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels. Documentation of the monitoring program failure criteria and action level targets will be contained in a documented evaluation. It is anticipated that the availability and reliability criterion for High Safety Significant Performance Monitoring Groups will use the guidance included in several industry documents tempered by site-specific operating experience, Fire PRA assumptions, and equipment types (and vendor data or valid design input when available). Industry documents such as the EPRI Fire Protection Equipment Surveillance Optimization and Maintenance Guide TR-1006756, Final Report July 2003, NFPA codes, and/or the NRC Fire Protection Significance Determination Process in addition to site specific operating experience data may be used. The monitoring program failure criteria and action level targets will be documented in the NFPA 805 Monitoring Program scoping document. Note that fire protection systems and features, NSCA equipment, SSCs required to meet the radioactive release criteria, and fire protection program elements that do not meet the screening criteria in Phase 2 will be included in the existing inspection and test programs and the system and program health programs. Reliability and availability criteria will not be assigned. Phase 4 - Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established. Monitoring consists of periodically gathering, trending, and evaluating information pertinent to the performance, and/or availability of the equipment and comparing the results with the established goals and performance criteria to verify that the goals are being met. Results of monitoring activities will be analyzed in a timely manner to assure that appropriate action is taken. The corrective action process will be used to address performance of fire protection and nuclear safety SSCs that do not meet performance criteria. Page 61 BSEP BSEP LAR Rev 2 LAR Rev 2 Page 61

CP&L 4.0 Compliance with NFPA 805 Requirements For fire protection systems and features and NSCA HSS equipment that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, corrective action in accordance with procedure, CAP-NGGC-0200 will be initiated to identify the negative trend. A corrective action plan will then be developed to ensure the performance returns to the established level. When applicable, a sensitivity study can be performed to determine the margin below the action level that still provides acceptable fire PRA results to help prioritize corrective actions if the action level is reached. A periodic assessment will be performed (i.e., at a frequency of approximately every two to three operating cycles), taking into account, where practical, industry wide operating experience. Issues that will be addressed include:

      " Review systems with performance criteria. Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the fire protection and NSCA systems?
      " Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection and NSCA SSCs, programmatic elements and/ or functions need to be in scope?
      " Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?

! Page 62 I IBSEP BSEPLARRev2 LAR Rev 2 Page 62

CP&L 4.0 Compliance with NFPA 805 Requirements CP&L 4.0 Compliance with NFPA 805 Requirements Fully describe process used* Figure 4 NFPA 805 Monitoring - Scoping and Screening 4.7 Program Documentation, Configuration Control, and Quality Assurance 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805 In accordance with the requirements and guidance in NFPA 805 Section 2.7.1, and NEI 04-02, BSEP has documented analyses to support compliance with l I BSEP LAR Rev 2 Page 63

CP&L 4.0 Compliance with NFPA 805 Requirements 10 CFR 50.48(c). The analyses are being performed in accordance with CP&L's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analyses. Analyses, as defined by NFPA 805 Section 2.4, performed to demonstrate compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy. Note these analyses do not include items such as periodic tests, hot work permits, fire impairments, etc. The Fire Protection Design Basis Document described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 have been created as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation. Figure 4-9 shows the Planned Post-Transition Documents. I Page 64 I IBSEP BSEPLARRev2 LAR Rev 2 Page 64

CP&L 4.0 Compliance with NFPA 805 Requirements NFPA 805 DOCUMENTS NSCA Database NSEL Comp I Cables [pRAEquipmentPRA Eqip~on I[Nnp~ Non-Power and Data Equipment and Data NSCA CALCULATION Comp & Cable FA Assessment Method/Results Method/Results Revised License Condition Treatmentsd OMA SSA Drawings NSCA SUPPORTING INFO Revised UFSAR Manual Action T-H Calculations Feasibility [ B-2 Table [i B-iabýe FIRE SAFETY ANALYSIS (DBD) Coordination Plant DBDs that " On a Fire Area Basis Calculations / Support NSCA - Fire Area Description MHIF

• FHA Database information
• Nuclear Safety Performance Criteria Compliance Summary (NEI 04-02 B-3 Table S ,,......................................................... . ...

Results)

• Non-Power Evaluation Results Summary Non-Power Mode NSCA Treatment - Radioactive Release Summary

                                                                                        " On a Generic Basis Non-Power Operations Calculations                                       - B-1 Table Results
                                                                                          - Radioactive Release (Training)
                                                                                          - Monitoring Program NFPA 805 FIRE RISK EVALUATIONS Fire Risk Evaluation Calculation(s)

S i................ -...............................................

        ------------------------------------------                 ------ ----------

0. Fire PRA FHA DATABASE DATA Ignition Sources FP Systems and

                    & Scenarios                          Features Data I                Inventory of                             B-1 Table Hazards                           Detailed Data S....... .....................        .........................................

FHA SUPPORT DOCUMENTATION FP Systems Code Compliance FP Drawings Evaluations Bold text indicates new NFPA 805 documents Engineering FP System and Engineeng

            =: FeatureFeatre    DBDs DI~s                       Equivalency Evlaon i ,.                                       Evaluations Radioactive Fire Pre-Plans                       Release Review S i.......... ........ .. . .. . . . .. . . . . . Calculation
                                                            . .. .. .. .. . . . .

Figure 4 NFPA 805 Planned Post-Transition Documents and Relationships l Page 65 I I BSEP BSEPLARRev2 LAR Rev 2 Page 65

CP&L 4.0 Compliance with NFPA 805 Requirements 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805 Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) is subject to CP&L configuration control processes that meet the requirements of Section 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes impacting the fire protection program are reviewed appropriately. The RI-PB post transition change process methodology is based upon the requirements of NFPA 805, and industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-2. Table 4-2 Change Evaluation Guidance Summary Table Document Section(s) Topic NFPA 805 2.2(h), 2.2.9, 2.4.4, A.2.2(h), A.2.4.4, Change Evaluation D.5 NEI 04-02 5.3, Appendix B, Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (Appendix I) RG 1.205 C.2.2.4, C.3.1, C.3.2, C.4.3 Risk Evaluation, Standard License Condition, Change Evaluation Process, Fire PRA The Plant Change Evaluation Process consists of the following 4 steps and is depicted in Figure 4-10:

   " Defining the Change
   " Performing the Preliminary Risk Screening.
   " Performing the Risk Evaluation
   " Evaluating the Acceptance Criteria Change Definition The Change Evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).

1. The baseline is defined as that plant condition or configuration that is consistent with the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).

2. The changed or altered condition or configuration that is not consistent with the Design Basis and Licensing Basis is defined as the proposed alternative.

Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes). BSEP LAR Rev 2 Page 66

CP&L 4.0 Compliance with NFPA 805 Requirements The characteristics of an acceptable screening process that meets the "assessment of the acceptability of risk" requirement of Section 2.4.4 of NFPA 805 are:

      " The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk.
      " The screening process must be documented and be available for inspection by the NRC.
      " The screening process does not pose undue evaluation or maintenance burden.

If any of the above is not met, proceed to the Risk Evaluation step. Risk Evaluation The screening is followed by engineering evaluations. The results of these evaluations are then compared to the acceptance criteria. Changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition. The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature. The risk evaluation involves the application of risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below. Acceptability Determination The Change Evaluations are assessed for acceptability using the ACDF (i.e., change in core damage frequency) and ALERF (i.e., change in large early release frequency) criteria from the license condition. The proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained. ! Page 67 I IBSEP BSEPLARRev2 LAR Rev 2 Page 67

CP&L 4.0 Compliance with NFPA 805 Requirements Defining the Change (5.3.2) License No VtChp3r Amendment Requestprvosyarve Yes License Amendment Request NOT Required Preliminary Risk Screening (5.3.3) Risk Evaluation (5.3.4) PRA Capability Category Assessment Fire PRA Capability Categorytr Assessment Acceptance Criteria (5.3.5) No Figure 4-10 Plant Change Evaluation [NEI 04-02 Figure 5-1] Note references in Figure refer to NEI 04-02 Sections I Page 68 IBSEP BSEPLARRev2 LAR Rev 2 Page 68

CP&L 4.0 Compliance with NFPA 805 Requirements The BSEP Fire Protection Program configuration is defined by the program documentation. The existing configuration control processes for modifications, calculations and analyses, and Fire Protection Program License Basis Reviews will be utilized to maintain configuration control of the Fire Protection program documents. The configuration control procedures which govern the various BSEP documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements (Implementation Item in Attachment S). Several NFPA 805 document types, such as NSCA Supporting Information, Non-Power Mode NSCA Treatment, generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. The new procedures will be modeled after the existing processes for similar types of documents and databases. System level design basis documents will be revised to reflect the NFPA 805 role that the system components now play. The process for capturing the impact of proposed changes to the plant on the Fire Protection Program will continue to be a multiple step review. The first step of the review is an initial screening for process users to determine if there is a potential to impact the Fire Protection program as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used. Reviews that identify potential Fire Protection program impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, Fire PRA) to ascertain the program impacts, if any. If Fire Protection program impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:

      " Deterministic Approach: Comply with NFPA 805, Chapter 3 and 4.2.3 requirements
      " Performance-Based Approach: Utilize the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process would be used to determine if the proposed change could be implemented "as-is" or whether prior NRC approval of the proposed change is required.

This process follows the requirements in NFPA 805 and the guidance outlined in RG 1.174, which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. The plant documents that ensure these requirements are met are: CAP-NGGC-0200 - Condition Identification and Screening Process EGR-NGGC-0005 - Engineering Change ESGO101 N - Safe Shutdown Engineer (Post-NFPA 805 Transition) ESGO102N - Fire Protection Plant Change Impact Review ESGO1 03N - Circuit Analysis (Post-NFPA 805 Transition) ! I BSEP LAR Rev 2 Page 69

CP&L 4.0 Compliance with NFPA 805 Requirements ESGO104N - Fire Protection Engineer (Post-NFPA 805 Transition) ESG0105N - Basic Fire Modeling 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality CP&L will maintain the existing fire protection quality assurance program. During the transition to 10 CFR 50.48(c), BSEP performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Fire PRA Quality Configuration control of the Fire PRA model will be maintained by integrating the Fire PRA model into the existing processes used to ensure configuration control of the internal events PRA model. This process complies with Section 1-5 of the ASME PRA Standard and ensures that CP&L maintains an as-built, as-operated PRA model of the plant. The process has been peer reviewed. Quality assurance of the Fire PRA is assured via the same processes applied to the internal events model. This process follows the guidance outlined in RG 1.174, which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Although the entire scope of the formal 10 CFR 50, Appendix B program is not applied to the PRA models or processes in general, often parts of the program are applied as a convenient method of complying with the requirements of RG 1.174. For instance, the procedure which addresses independent review of calculations for 10 CFR 50, Appendix B, is applied to the PRA model calculations, as well. With respect to Quality Assurance Program requirements for independent reviews of calculations and evaluations, those existing requirements for Fire Protection Program documents will remain unchanged. CP&L specifically requires that the calculations and evaluations in support of the NFPA 805 LAR, exclusive of the Fire PRA, be performed within the scope of the QA program which requires independent review as defined by plant procedures. As recommended by NUREG/CR-6850, the sources of uncertainty in the Fire PRA were identified and specific parameters were analyzed for sensitivity in support of the NFPA 805 Fire Risk Evaluation process. Specifically with regard to uncertainty, an uncertainty and sensitivity matrix was developed and included with BNP-PSA-080. In addition, sensitivity to uncertainty associated with specific Fire PRA parameters was quantitatively addressed in BNP-PSA-095. While the removal of conservatism inherent in the Fire PRA is a long-term goal, the Fire PRA results were deemed sufficient for evaluating the risk associated with this application. While CP&L continues to strive toward a more "realistic" estimate of fire risk, use of mean values continues to be the best estimate of fire risk. During the Fire Risk Evaluation process, the uncertainty and sensitivity associated with specific Fire PRA parameters were considerations in the evaluation of the change in risk relative to the applicable acceptance thresholds. ! I BSEP LAR Rev 2 Page 70

CP&L 4.0 Compliance with NFPA 805 Requirements Specific Requirements of NFPA 805 Section 2.7.3 NFPA 805 Section 2.7.3.1 - Review Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with procedures that require independent review. Reference plant procedures: EGR-NGGC-0003 - Design Review Requirements EGR-NGGC-0005 - Engineering Change EGR-NGGC-0017 - Preparation and Control of Design Analyses and Calculations NFPA 805 Section 2.7.3.2 - Verification and Validation 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. NFPA 805 Section 2.7.3.3 - Limitations of Use Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were applied appropriately as required by Section 2.7.3.3 of NFPA 805. NFPA 805 Section 2.7.3.4 - Qualification of Users Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805. During the transition to 10 CFR 50.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g., fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805, Section 2.7.3.4. Post-transition, for personnel performing fire modeling or Fire PRA development and evaluation, CP&L has developed and maintains qualification requirements for individuals assigned various tasks. Position-Specific Guides have been 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. The following Training Guides have been developed and implemented. ESGO089N - Fire Probabilistic Safety Assessment Engineer (Quantification), ESGO093N - Fire Probabilistic Safety Assessment Engineer (Initial Development), and ESGO094N - Fire Probabilistic Safety Assessment Engineer (Data Development), and ESGO105N - Basic Fire Modeling NFPA 805 Section 2.7.3.5 - Uncertainty Analysis 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 ! I BSEP LAR Rev 2 Page 71

CP&L 4.0 Compliance with NFPA 805 Requirements fire modeling and Fire PRA development. Note: 10 CFR 50.48(c)(2)(iv) states that NFPA 805, Section 2.7.3.5 is not required for the deterministic approach because conservatism is included in the deterministic criteria. 4.8 Summary of Results 4.8.1 Results of the Fire Area Review A summary of the NFPA 805 compliance basis and the required fire protection systems and features is provided in Attachment C. The table provides the following information from the NEI 04-02 Table B-3:

    " Fire Area / Fire Zone: Fire Area/Zone Identifier.
    "

Description:

Fire Area/Zone Description.

    " NFPA 805 Regulatory Basis: Post-transition NFPA 805 Chapter 4 compliance basis
    " Required Fire Protection System / Feature: Detection / suppression required in the Fire Area based on NFPA 805 Chapter 4 compliance. Other Required Features may include Electrical Raceway Fire Barrier Systems, fire barriers, etc.

The documentation of required fire protection systems and features does not include the documentation of the fire area boundaries. Fire area boundaries are required and documentation of the fire area boundaries has been performed as part of reviews of engineering evaluations, licensing actions, or as part of the reviews of the NEI 04-02 Table B-1 process. The basis for the requirement of the fire protection system / feature is designated as follows: o S - Separation Criteria: Systems/Features required for Chapter 4 Separation Criteria in Section 4.2.3 o E - EEEE/LA Criteria: Systems/Features required for acceptability of Existing Engineering Equivalency Evaluations / NRC approved Licensing Action (i.e., Exemptions/Deviations/Safety Evaluations) (Section 2.2.7) o R - Risk Criteria: Systems/Features required to meet the Risk Criteria for the Performance-Based Approach (Section 4.2.4) o D - Defense-in-depth Criteria: Systems/Features required to maintain adequate balance of Defense-in-Depth for a Performance-Based Approach (Section 4.2.4) An evaluation of DID was performed for all fire areas as detailed in project procedure FPIP-1129, NFPA 805 Fire Safety Analysis. This evaluation was performed for all areas, regardless of whether NFPA 805 compliance was demonstrated using a performance based approach or a deterministic approach. Although a discussion of DID features is not strictly required for areas that are deterministically compliant, the decision to include the evaluation for such areas was based on two factors. First, it was seen as a way of enhancing the overall approach to providing the plant's desired level of fire protection to that area. Second, if future changes to deterministic areas dictate that a performance based approach is desired, then including these features as credited DID features now will facilitate that transition. The regulatory basis for each fire area is provided in BSEP LAR Rev 2 Page 72

CP&L 4.0 Compliance with NFPA 805 Requirements Attachment C, but the presence of deterministic features in the DID discussion does not alter any conclusions regarding the post transition licensing basis. Attachment W contains the results of the Fire Risk Evaluations, additional risk of recovery actions, and the change in risk on a fire area basis. 4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase Planned modifications, studies, and evaluations to comply with NFPA 805 are described in Attachment S. In Attachment S, two tables are listed. Table S-1 identifies Plant Modifications required to be completed. Table S-2 identifies training, programs, personnel equipment, and document changes and upgrades required to be completed. The Fire PRA model will represent the as-built, as-operated and maintained plant following completion of the risk related modifications identified in Attachment S. In the event the PRA model requires revision following completion of the modifications and implementation items listed in Attachment S, the changes will be controlled through normal BSEP processes. These changes are not expected to be significant. The Main Control Room ceiling modification is the only outstanding change with respect to its inclusion in the Fire PRA model. 4.8.3 Supplemental Information -Other Licensee Specific Issues The development of a FPRA requires that assumptions and methods be expanded and updated to provide more realistic treatment of the data and the phenomena involved. The updates and expansion of methods introduce differences in plant specific results depending on which alternatives are used. This section captures the sensitivities and insights based on these alternatives. These alternatives may be based new analysis methods, new data, or deviations from guidance in NUREG/CR-6850 which would require sensitivity analyses to be performed for the license application. 4.8.3.1 Unreviewed Analysis Methods The peer review of the Brunswick plants' fire PRA identified one method that had not been reviewed by the methods panel concerning the use of a split fraction for closed cabinet fires that result in damage outside of the cabinet. This method was reviewed by the NRC for the Harris plant NFPA 805 pilot effort as documented in Section 3.4.7 and Table 3.4-6 of the Safety Evaluation for the Harris Plant license amendment (ML101130535). There is variation in the methods in treating how MCCs can be treated as "closed" cabinets. If a cabinet were always "closed" there would be no fire impact on external targets. However, there is always the potential for the cabinet to already be open or an arc fault to have enough energy to open the cabinet. For the Brunswick FPRA, it was assumed that one out of ten MCC fires may result in an "open" cabinet configuration. I Page 73 II BSEP BSEPLARRev2 LAR Rev 2 Page 73

CP&L 4.0 Compliance with NFPA 805 Requirements This is not applied to the HRR as a severity factor, but as a split fraction on the likelihood that the cabinet remains "closed." Because the guidance for characterizing closed cabinets at the Brunswick plant was the same as that used for Harris Plant pilot effort, the use of split fractions as described above is acceptable. A sensitivity analysis was performed on this method for the Brunswick plant fire PRA. The sensitivity analysis essentially removed the split fraction, effectively treating the closed MCCs as always open. The results of the sensitivity for the "closed" cabinet method are provided below. ! Page 74 I IBSEP LAR Rev 2 BSEPLARRev2 Page 74

CP&L 4.0 Compliance with NFPA 805 Requirements Table 4 Closed MCCs Sensitivity Delta CDF and Delta LERF Results Unit I Unit 2 ACDF [/yr] ALERF [/yr] ACDF [/yr] ALERF [/yr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) 1.9E-06 3.8E-07 2.4E-06 4.3E-07 VFDRs (+2.2%) (0.0%) (0.0%) (0.0%) Recovery Actions 8.9E-07 8.9E-08 8.9E-07 8.9E-08 (0.0%) (0.0%) (0.0%) (0.0%) Total 2.8E-06 4.7E-07 3.3E-06 5.2E-07 (+1.5%) (0.0%) (0.0%) (0.0%) Differences in percentages are due to rounding. Table 4 Closed MCCs Sensitivity Total CDF and LERF Results Unit I Unit 2 CDF [/yr] LERF [lyr] CDF [lyr] LERF [lyr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) Internal Events plus 1.4E-05 6.2E-07 1.4E-05 6.2E-07 External Flooding and (0.0%) (0.0%) (0.0%) (0.0%) High Winds Fire[11 1.6E-05 3.9E-06 1.3E-05 1.4E-06 (+6.7%) (0.0%) (0.0%) (0.0%) Fire - Recovery Actions[2 1 1.OE-06 1E-07 1.OE-06 1E-07 (0.0%) (0.0%) (0.0%) (0.0%) Total 3.1 E-05 4.6E-06 2.8E-05 2.1 E-06 (+3.3%) (0.0%) (0.0%) (0.0%) Fire results do not credit control room abandonment for loss of control sequences. [2] Values are for recovery actions associated with control room abandonment due to environmental reasons. Differences in percentages are due to rounding. I Page 75 I I BSEP BSEPLARRev2 LAR Rev 2 Page 75

CP&L 4.0 Compliance with NFPA 805 Requirements 4.8.3.2 Concerns with NUREG/CR-6850 CPT Credit Based on preliminary results for fire circuit testing, the credit allowed in Tables 10-1 and 10-3 of NUREG/CR-6850 for Control Power Transformers (CPT) in AC circuits was questioned by NRR. This is based on an RAI letter to Duane Arnold (ML12031A112). The sensitivity analysis was performed by removing the approximately factor of two reduction in failure mode probability estimates between cables with CPT and those without. The results of the sensitivity analysis for the CPT credit are provided below. Table 4 CPT Sensitivity Delta CDF and Delta LERF Results Unit I Unit 2 ACDF [lyr] ALERF [lyr] ACDF [lyr] ALERF [lyr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) 1.9E-06 3.8E-07 2.5E-06 4.4E-07 VFDRs (+1.6%) (+0.3%) (+4.6%) (+1.6%) Recovery Actions 8.9E-07 8.9E-08 8.9E-07 8.3E-08 (0.0%) (0.0%) (0.0%) (0.0%) 2.8E-06 4.7E-07 3.4E-06 5.3E-07 Total (+1.1%) (+0.2%) (+3.3%) (+1.3%) Differences inpercentages are due to rounding. Table 4 CPT Sensitivity Total CDF and LERF Unit I Unit 2 CDF [Iyr] LERF [/yr] CDF [lyr] LERF [Iyr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) Internal Events plus 1.4E-05 6.2E-07 1.4E-05 6.2E-07 External Flooding and (0.0%) (0.0%) (0.0%) (0.0%) High Winds Firell1 1.5E-05 4.OE-06 1.3E-05 1.4E-06 (0.0%) (+2.6%) (0.0%) (0.0%) Fire - Recovery Actions[2] 1.OE-06 1E-07 1.OE-06 1E-07 (0.0%) (0.0%) (0.0%) (0.0%) Total 3.OE-05 4.7E-06 2.8E-05 2.1 E-06 (0.0%) (+2.2%) (0.0%) (0.0%) [M] Fire results do not credit control room abandonment for loss of control sequences. [2] Values are for recovery actions associated with control room abandonment due to environmental reasons. Differences in percentages are due to rounding. ! I BSEP LAR Rev 2 Page 76

CP&L 4.0 Compliance with NFPA 805 Reauirements 4.8.3.3 Sensitivity Analysis Required by FAQ 08-0048 In order to use the updated fire bin ignition frequencies provided in Supplement 1 to NUREG/CR-6850, a sensitivity analysis must be performed comparing the impact of those bins characterized by an alpha from the EPRI TR-1 016735 analysis that is less than or equal to 1. While the new point estimates for the bin ignition frequencies better represent the data, uncertainties are large and a sensitivity analysis using the old frequencies was required to assess the potential impact of using the new frequencies. Since the largest contributor to delta fire risk for Brunswick is control room abandonment, the factor of three increase in Main Control Board bin ignition frequency results in large changes in risk metrics. This resulted in a doubling of the delta CDF and delta LERF metrics and large changes in total CDF and LERF. Explanation on why the impact of control room abandonment is conservative is provided in section 4.8.3.4. The increases in risk metrics are provided below. I Page 77 I IBSEP BSEPLARRev2 LAR Rev 2 Page 77

CP&L 4.0 Compliance with NFPA 805 Requirements Table 4 Ignition Frequency Sensitivity Delta CDF and Delta LERF Results Unit I Unit 2 ACDF [lyr] ALERF [/yr] ACDF [/yr] ALERF [Iyr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) 3.3E-06 7.5E-07 3.1E-06 8.4E-07 VFDRs (+80%) (+98%) (+27%) (+94%) Recovery Actions 2.7E-06 2.7E-07 2.7E-06 2.7E-07 (+197%) (+197%) (+197%) (+197%) Total 6.OE-06 1.OE-06 5.7E-06 1.1 E-06 (+118%) (+117%) (+73%) (+111%) Differences inpercentages are due to rounding. Table 4 Ignition Frequency Sensitivity Total CDF and LERF Unit I Unit 2 CDF [/yr] LERF [/yr] CDF [/yr] LERF [lyr] (Change from (Change from (Change from (Change from baseline) baseline) baseline) baseline) Internal Events plus 1.4E-05 6.2E-07 1.4E-05 6.2E-07 External Flooding and (0.0%) (0.0%) (0.0%) (0.0%) High Winds Fire[1] 2.6E-05 7.5E-06 2.1 E-05 2.6E-06 (+73%) (+92%) (+62%) (+86%) Fire - Recovery Actions[2] 3.1 E-06 3.1 E-07 3.1 E-06 3.1 E-07 (+210%) (+210%) (+210%) (+210%) Total 4.3E-05 8.4E-06 3.8E-05 3.5E-06 (+44%) (+82%) (+36%) (+67%) Fire results do not credit control room abandonment for loss of control sequences. [2] Values are for recovery actions associated with control room abandonment due to environmental reasons. Differences in percentages are due to rounding. I Page 78 IBSEPLARRev2 I BSEP LAR Rev 2 Page 78

CP&L 4.0 Compliance with NFPA 805 Requirements 4.8.3.4 Main Control Room Abandonment The control room abandonment has a detailed human error analysis. Control abandonment was not credited for loss of control scenarios. The Brunswick control room has two main areas that contribute to control room abandonment: the area where in the control room staff generally manipulates controls and, the area that is outside that region. The region that the control room staff generally occupies while manipulating the controls has a much lower ceiling height and smaller footprint than the remaining area and, as such, has a much shorter time to suppress a fire prior to reaching control room habitability concerns that require abandonment. A shorter time to suppress the fire increases the frequency of control room abandonment which is directly proportional to CDF and LERF contributions for control room abandonment. The control room abandonment risk is conservative since the fire frequency contribution from the cabinets that comprise the boundary for the control manipulation area is all applied to the small area when there is significant probability that the fire would vent out of the back of the panels to the larger area with higher ceilings. Since methods to determine a split fraction of fires that vent to the rear of the panel were not peer reviewed, all the frequency was conservatively applied to the smaller region with the low ceiling resulting in conservative times to conditions requiring control room abandonment. 4.8.3.5 Reduction in Transient Source Heat Release Rate Following transition to NFPA 805, BSEP will adopt a more restrictive transient control program that will nominally limit the transient fire HRR to the 143 kW range instead of the 317 kW range. The transient control program is the fleet program and is already in use at HNP. The 143 kW range was used for the transient fire locations in all areas except for the turbine building, which uses a 317 kW HRR. 4.8.3.6 Incipient Detection in Main Control Boards The FPRA credits the use of air-aspirated incipient detection, also known as Very Early Warning Fire Detection Systems (VEWFDS) in NFPA 76, in the Main Control Boards (MCBs) because that modification is expected to be completed prior to the transition to NFPA 805. To support the use of incipient detection, a walkdown was performed for a representative sample of BSEP MCBs and determined the fraction of fast-acting components to be very small (less than 0.5%) of the total component count. A two-part sensitivity analysis was performed using the currently installed in-panel ion smoke detection rather than the incipient detection. In the first part, the NUREG/CR-6850 Appendix L method was used to determine the frequency of self fires that cause fire damage only within the MCBs. In the second part, the ignition frequency of NUREG/CR-6850 Supplement 1 was modified by the "normal" non-suppression probability for ion smoke detectors and manual detection/suppression for fires that also cause damage in the zone-of-influence outside the MCBs. For the zone-of-influence I I BSEP LAR Rev 2 Page 79

CP&L 4.0 Compliance with NFPA 805 Requirements fires, the human reliability analysis for LERF was further refined to account for operator actions to secure ac and dc power to primary containment isolation valves during MCR abandonment not related to habitability issues. Since these operator actions are performed either in the control room or in-transient to support the remote shutdown panels, they are not considered Recovery Actions for the purpose of evaluating the VFDRs. Table 4 Incipient Detection Sensitivity Delta CDF and Delta LERF Results Unit I ACDF [lyr] ALERF [lyr] (Change from (Change from baseline) baseline) 1.86E-06 6.03E-07 VFDRs (0%) (+59%) Recovery Actions[1] 8.93E-07 8.93E-08 2.75E-06 6.92E-07 Total (0%) (+48%) [1] Values are for recovery actions associated with control room abandonment due to environmental reasons and address those actions away from the remote shutdown panel. Table 4 Incipient Detection Sensitivity Total CDF and LERF Results Unit I CDF [/yr] LERF [Iyr] (Change from (Change from baseline) baseline) Internal Events plus External Flooding and 1.4E-05 6.2E-07 High Winds 1.9E-05 4.2E-06 Fire[l] (+27%) (+8%) Fire - Recovery Actions[2] 1.0E-06 1.0E-07 3.4E-05 4.9E-06 Total (+13%) (+7%) [1] Fire results credit operator actions to secure power during control room abandonment for loss of control sequences. [2] Values are for recovery actions associated with control room abandonment due to environmental reasons. These results are considered conservative because this sensitivity analysis took no credit for more realistic detailed fire modeling (e.g., shielded targets) for targets above the MCBs. BSEP LAR Rev 2 Page 80

CP&L 4.0 Compliance with NFPA 805 Requirements The sensitivity analysis was performed for Unit 1, but the risk insights are applicable to Unit 2 because the panels of interest are similar with regard to the arrangement of instrumentation and controls, detection and suppression, and operator actions. ! Page 81 I BSEPLARRev2 BSEP LAR Rev 2 Page 81

CP&L 5.0 Regulatory Evaluation

5.0 REGULATORY EVALUATION

5.1 Introduction - 10 CFR 50.48 On July 16, 2004, the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements. 10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition (NFPA 805), as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning. The voluntary adoption of 10 CFR 50.48(c) by BSEP does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, GDC 3, Fire Protection. The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference Federal Register (FR) Notice 69 FR 33536 dated June 16, 2004, ML041340086).

     "NFPA 805 does not supersede the requirementsof GDC 3, 10 CFR 50.48(a), or 10 CFR 50.48(0. Those regulatory requirementscontinue to apply to licensees that adopt NFPA 805. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirements may be met is different than under 10 CFR 50.48(b). Specifically, whereas GDC 3 refers to SSCs importantto safety, NFPA 805 identifies fire protection systems and features required to meet the Chapter I performance criteria through the methodology in Chapter4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii)requirementto limit fire damage to SSCs important to safety so that the capabilityto safely shut down the plant is ensured is satisfied by meeting the performance criteria in Section 1.5.1 of NFPA 805. The Section 1.5.1 criteria include provisions for ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries,and process monitoring are achieved and maintained.

This methodology specifies a process to identify the fire protection systems and features requiredto achieve the nuclear safety performance criteria in Section 1.5 of NFPA 805. Once a determinationhas been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design must meet any applicablerequirementsof NFPA 805, Chapter3. Having identified the requiredfire protection systems and features, the licensee selects either a deterministic or performance-basedapproach to demonstrate that the performance criteria are satisfied. This process satisfies the GDC 3 requirement to design and locate SSCs importantto safety to minimize the probabilityand effects of fires and explosions." (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086) The new rule provides actions that may be taken to establish compliance with 10 CFR 50.48(a), which requires each operating nuclear power plant to have a fire protection program plan that satisfies GDC 3, as well as specific requirements in that section. The transition process described in 10 CFR 50.48(c)(3)(ii) provides, in pertinent parts, that a licensee intending to adopt the new rule must, among other things, "modify the fire protection plan required by paragraph (a) of that section to reflect the licensee's decision to comply with NFPA 805." Therefore, to the extent that the I I BSEP LAR Rev 2 Page 82

CP&L 5.0 Regulatory Evaluation contents of the existing fire protection program plan required by 10 CFR 50.48(a) are inconsistent with NFPA 805, the fire protection program plan must be modified to achieve compliance with the requirements in NFPA 805. All other requirements of 10 CFR 50.48(a) and GDC 3 have corresponding requirements in NFPA 805. A comparison of the current requirements in Appendix R with the comparable requirements in Section 3 of NFPA 805 shows that the two sets of requirements are consistent in many respects. This was further clarified in FAQ 07-0032, 10 CFR 50.48(a), and GDC 3 clarification (ML081400292). The following tables provide a cross-reference of fire protection regulations associated with the post-transition BSEP fire protection program and applicable industry and BSEP documents that address the topic. 10 CFR 50.48(a) Table 5-1 10 CFR 50.48(a) - ApplicabilitylCompliance Reference 10 CFR 50.48(a) Section(s) ApplicabilitylCompliance Reference (1) Each holder of an operating license issued under this See below part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part. This fire protection plan must: (i) Describe the overall fire protection program for the NFPA 805 Section 3.2 facility; NEI 04-02 Table B-1 (ii) Identify the various positions within the licensee's NFPA 805 Section 3.2.2 organization that are responsible for the program; NEI 04-02 Table B-1 (iii) State the authorities that are delegated to each of NFPA 805 Section 3.2.2 these positions to implement those responsibilities; and NEI 04-02 Table B-1 (iv) Outline the plans for fire protection, fire detection NFPA 805 Section 2.7 and Chapters 3 and 4 and suppression capability, and limitation of fire NEI 04-02 B-1 and B-3 Tables damage. (2) The plan must also describe specific features See below necessary to implement the program described in paragraph (a)(1) of this section such as: (i) Administrative controls and personnel requirements NFPA 805 Sections 3.3.1 and 3.4 for fire prevention and manual fire suppression NEI 04-02 Table B-1 activities; (ii)Automatic and manually operated fire detection and NFPA 805 Sections 3.5 through 3.10 and suppression systems; and Chapter 4 NEI 04-02 B-1 and B-3 Tables (iii) The means to limit fire damage to structures, NFPA 805 Section 3.3 and Chapter 4 systems, or components important to safety so that the NEI 04-02 B-3 Table capability to shut down the plant safely is ensured. (3) The licensee shall retain the fire protection plan and NFPA 805 Section 2.7.1.1 requires that each change to the plan as a record until the documentation (Analyses, as defined by NFPA 805 Commission terminates the reactor license. The 2.4, performed to demonstrate compliance with this licensee shall retain each superseded revision of the standard) be maintained for the life of the plant. procedures for 3 years from the date it was RDC-NGGC-0001 superseded. I Page 83 I IBSEP LAR Rev 2 Page 83

CP&L 5.0 Regulatory Evaluation Table 5-1 10 CFR 50.48(a)- Applicability/Compliance Reference 10 CFR 50.48(a) Section(s) Applicability/Complicance Reference (4) Each applicant for a design approval, design Not applicable. BSEP is licensed under certification, or manufacturing license under part 52 of 10 CFR 50. this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of appendix A to this part. General Design Criterion 3 Table 5-2 GDC 3 - Applicability/Compliance Reference GDC 3, Fire Protection, Statement Applicability/Compliance Reference Structures, systems, and components important to NFPA 805 Chapters 3 and 4 safety shall be designed and located to minimize, NEI 04-02 B-1 and B-3 Tables consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be NFPA 805 Sections 3.3.2, 3.3.3, 3.3.4, 3.11.4 used wherever practical throughout the unit, NEI 04-02 B-1 Table particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate NFPA 805 Chapters 3 and 4 capacity and capability shall be provided and designed NEI 04-02 B-1 and B-3 Tables to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that NFPA 805 Sections 3.4 through 3.10 and 4.2.1 their rupture or inadvertent operation does not NEI 04-02 Table B-3 significantly impair the safety capability of these structures, systems, and components I Page 84 I I BSEP LAIR Rev 2 BSEPLARRev2 Page 84

CP&L 5.0 Regulatory Evaluation 10 CFR 50.48(c) Table 5-3 10 CFR 50.48(c) - ApplicabilitylCompliance Reference 10 CFR 50.48(c) Section(s) Applicability/Compliance Reference (1) Approval of incorporationby reference. National Fire Protection Association General Information. (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for NFPA 805 (2001 edition) is Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), the edition used. which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51. (2) Exceptions, modifications, and supplementation of NFPA 805. As used in General Information. this section, references to NFPA 805 are to the 2001 Edition, with the NFPA 805 (2001 edition) is following exceptions, modifications, and supplementation: the edition used. (i) Life Safety Goal, Objectives, and Criteria. The Life Safety Goal, The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed. Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR. (ii) Plant Damage/Business InterruptionGoal, Objectives, and Criteria.The The Plant Damage/Business Plant Damage/Business Interruption Goal, Objectives, and Criteria of Interruption Goal, Objectives, Chapter 1 are not endorsed. and Criteria of Chapter 1 of NFPA 805 are not part of the LAR. (iii) Use of feed-and-bleed. In demonstrating compliance with the BSEP is a BWR. This is not performance criteria of Sections 1.5.1 (b) and (c), a high-pressure applicable. charging/injection pump coupled with the pressurizer power-operated relief valves (PORVs) as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for pressurized-water reactors (PWRs) is not permitted. (iv) Uncertainty analysis. An uncertainty analysis performed in accordance Uncertainty analysis was not with Section 2.7.3.5 is not required to support deterministic approach performed for deterministic calculations. methodology. (v) Existing cables. In lieu of installing cables meeting flame propagation Electrical cable construction tests as required by Section 3.3.5.3, a flame-retardant coating may be complies with a flame applied to the electric cables, or an automatic fixed fire suppression system propagation test that was may be installed to provide an equivalent level of protection. In addition, the found acceptable to the NRC italicized exception to Section 3.3.5.3 is not endorsed, as documented in NEI 04-02 Table B-i. (vi) Water supply and distribution. The italicized exception to Section 3.6.4 is BSEP complies as not endorsed. Licensees who wish to use the exception to Section 3.6.4 documented in Attachment A. must submit a request for a license amendment in accordance with See NEI 04-02 Table B-i. paragraph (c)(2)(vii) of this section. I Page 85 I I BSEP LAR Rev 2 BSEPLARRev2 Page 85

CP&L 5.0 Regulatory Evaluation Table 5-3 10 CFR 50.48(c) - ApplicabilitylCompliance Reference 10 CFR 50.48(c) Section(s) ApplicabilitylCompliance Reference (vii) Performance-based methods. Notwithstanding the prohibition in Section The use of performance-3.1 against the use of performance-based methods, the fire protection based methods for NFPA 805 program elements and minimum design requirements of Chapter 3 may be Chapter 3 is requested. See subject to the performance-based methods permitted elsewhere in the Attachment L. standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). (3) Compliance with NFPA 805. See below (i) A licensee may maintain a fire protection program that complies with The LAR was submitted in NFPA 805 as an alternative to complying with paragraph (b) of this section accordance with for plants licensed to operate before January 1, 1979, or the fire protection 10 CFR 50.90. The LAR license conditions for plants licensed to operate after January 1, 1979. The included applicable license licensee shall submit a request to comply with NFPA 805 in the form of an conditions, orders, technical application for license amendment under § 50.90. The application must specifications/bases that identify any orders and license conditions that must be revised or needed to be revised and/or superseded, and contain any necessary revisions to the plant's technical superseded. specifications and the bases thereof. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate. Any approval by the Director or the designee must be in the form of a license amendment approving the use of NFPA 805 together with any necessary revisions to the technical specifications. (ii) The licensee shall complete its implementation of the methodology in The LAR and transition report Chapter 2 of NFPA 805 (including all required evaluations and analyses) summarize the evaluations and, upon completion, modify the fire protection plan required by paragraph and analyses performed in (a) of this section to reflect the licensee's decision to comply with NFPA 805, accordance with Chapter 2 of before changing its fire protection program or nuclear power plant as NFPA 805. permitted by NFPA 805. (4) Risk-informed or performance-based alternatives to compliance with NFPA No risk-informed or 805. A licensee may submit a request to use risk-informed or performance- performance-based based alternatives to compliance with NFPA 805. The request must be in alternatives to compliance the form of an application for license amendment under § 50.90 of this with NFPA 805 (per chapter. The Director of the Office of Nuclear Reactor Regulation, or 10 CFR 50.48(c)(4)) were designee of the Director, may approve the application if the Director or utilized. See Attachment P. designee determines that the proposed alternatives: (i) Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). ! Page 86 I I BSEP LAR Rev 2 BSEPLARRev2 Page 86

CP&L 5.0 Regulatory Evaluation 5.2 Regulatory Topics 5.2.1 License Condition Changes The current BSEP fire protection license condition 2.B.(6) is being replaced with the standard license condition based upon Regulatory Position 3.1 of RG 1.205, as shown in Attachment M. 5.2.2 Technical Specifications BSEP conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded. BSEP determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the BSEP adoption of the new fire protection licensing basis. 5.2.3 Orders and Exemptions A review was conducted of the BSEP docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained. A discussion of affected orders and exemptions is included in Attachment 0. 5.3 Regulatory Evaluations 5.3.1 No Significant Hazards Consideration A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

     " Involve a significant increase in the probability or consequences of an accident previously evaluated; or
     " Create the possibility of a new or different kind of accident from any accident previously evaluated; or
     " Involve a significant reduction in a margin of safety.

This evaluation is contained in Attachment Q. Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. BSEP has evaluated the proposed amendment and determined that it involves no significant hazards consideration. 5.3.2 Environmental Consideration Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in I BSEP LAR Rev 2 Page 87

CP&L 5.0 Regulatory Evaluation 10 CFR 51.22(c). That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement. 5.4 Revision to the UFSAR After the approval of the LAR, in accordance with 10 CFR 50.71(e), the BSEP UFSAR will be revised. The content will be consistent with NEI 04-02. 5.5 Transition Implementation Schedule The following schedule for transitioning BSEP to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule:

       " Implementation of new NFPA 805 fire protection program to include procedure changes, process updates, and training to affected plant personnel. This will occur 180 days after NRC approval. If the turnover is due to fall within an outage window then the changes will be implemented 60 days after startup from the scheduled outage.
       " Modifications will be completed by the startup of the second refueling outage for each unit after issuance of the Safety Evaluation (SE). Appropriate compensatory measures will be maintained until modifications are complete.

I Page 88 I IBSEP BSEPLARRev2 LAR Rev 2 Page 88

CP&L 6.0 References

6.0 REFERENCES

The following references were used in the development of the TR. Additional references are in the Attachments. NRC Documents

1. Letter, NRC to NEI, Process for Frequently Asked Questions For Title 10 of The Code Of Federal Regulations, Part 50.48(c) Transitions, July 12, 2006 (ML061660105).

2. NRC Enforcement Policy, Policy Statement: Revision, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33684-33685.

3. NRC Generic Letter 86-10, Supplement 1, Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within the Same Fire Area, March 25, 1994.

4. NRC Regulatory Issue Summary 2007-19: Communicating Clarifications of Staff Positions in RG 1.205 Concerning Issues Identified During Pilot Application of NFPA Std 805, August 20, 2007 (ML071590227).

5. NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, April 2005.

6. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1 - November 2002.

7. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 - March 2009).

8. Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Revision 1, December 2009.

9. Voluntary Fire Protection Requirement for Light-Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative, Final Rule, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33536-33551.

Other Industry Documents

1. ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S 2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers/American Nuclear Society, New York, NY.

2. EPRI Fire Protection Equipment Surveillance Optimization and Maintenance Guide TR-1006756, Final Report July 2003

3. NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Revision 1, January 2005.

I I BSEP LAR Rev 2 Page 89

CP&L 6.0 References

4. NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Revision 2, May 2009.

5. NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c), Revision 2 April 2008.

6. NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition.

Licensee Correspondence

1. Letter CP&L to NRC, Letter of Intent to Transition to 10 CFR 40.48(c),

June 10, 2005 (ML051720404)

2. Letter NRC to CP&L, Grants Enforcement Discretion Regarding NFPA Standard 805, April 29, 2007 (ML070590625).

3. Letter NRC to CP&L, Issuance of Amendment Regarding Adoption of NFPA Standard 805, Safety Evaluation for the Shearon Harris Nuclear Power Plant, June 28, 2012 (ML1O01130535)

I BSEP LAR Rev 2 Page 90

Enclosure 4 Revised NFPA 805 Transition Report, Attachment A, NEI 04-02 Table B-i, Transition of FundamentalFire ProtectionProgramand Design Elements

CP&L Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements 84 Pages Attached Page A-I BSEP Rev 2 LAR Rev BSEP LAR 2 Page A-1

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.1 General Chapter 3 Requirement: 3.1* General. This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. These fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard. Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein. Compliance Statement Compliance Rasisq N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2 Fire Protection Plan Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.1 Intent Chapter 3 Requirement: 3.2.1 Intent. A site-wide fire protection plan shall be established. This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the plan's implementation. This section establishes the criteria for an integrated combination of components, procedures, and personnel to implement all fire protection program activities Compliance Statement Compliance Basis BSEP LAR Rev 2 Page A-2

CP&L Attachment A Complies No Additional Clarification. Reference Document DoDetals OAP-033,Fire Protection Program Manual All Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2 Management Policy Direction and Responsibility. Chapter 3 Requirement: 3.2.2* Management Policy Direction and Responsibility. A policy document shall be prepared that defines management authority and responsibilities and establishes the general policy for the site fire protection program. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details OAP-033,Fire Protection Program Manual Section 4 & Section 5.1.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.1 [Management Policy on Senior Management] Chapter 3 Requirement: 3.2.2.1* The policy document shall designate the senior management position with immediate authority and responsibility for the fire protection program. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details OAP-033,Fire Protection Program Manual Section 4.2.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.2 [Management Policy on Daily Administration] Chapter 3 Requirement: 3.2.2.2* BSEP LAR Rev 2 Page A-3

CP&L Attachment A The policy document shall designate a position responsible for the daily administration and coordination of the fire protection program and its implementation. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Docti talis OAP-033,Fire Protection Program Manual Section 4.2 & Figure 1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.3 [Management Policy on Interfaces] Chapter 3 Requirement: 3.2.2.3* The policy document shall define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In addition, this policy document shall identify the various plant positions having the authority for implementing the various areas of the fire protection program. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details OAP-033,Fire Protection Program Manual Section 1.1 & 5.1.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.4 [Management Policy on AHJ] Chapter 3 Requirement: 3.2.2.4* The policy document shall identify the appropriate AHJ for the various areas of the fire protection program. Compliance Statement Compliance Basis Complies with Clarification The Authority having Jurisdiction (AHJ) is understood to be the NRC or the utility itself. The AHJ for BNP is the NRC and the FPPM essentially accredits that accountability to the utility for all changes and additions where it can be shown that a decrease in the effectiveness of the BNP FP Program does not result. BSEP LAR Rev 2 Page A-4

CP&L Attachment A The NRC approved the licensing, construction, and operation of BNP and retains the right/responsibility for regulation, inspection, and audit of all BNP systems and facilities. As stated in Section 5.1.5 of the FPPM,

                                                "...changes can be identified by any individual utilizing the FPPM and submitted to the Fire Protection Program Manager for review and, as appropriate, incorporation into the FPPM."

Per Section 5.1.5 of the FPPM, "Changes to the content of the FPPM fall into one of two categories: those requiring prior NRC approval before implementation and those that may be implemented without prior NRC approval. A determination will be made by the Fire Protection Program Manager as to the impact the proposed change(s) has on the Fire Protection Program. If the proposed change(s) does not decrease the effectiveness of the program, the program may be revised and implemented in accordance with this FPPM. If the proposed change(s) adversely impacts the ability to achieve and maintain safe shutdown, approval must be obtained from the NRC prior to implementation." Reference Document DoDetals OAP-033,Fire Protection Program Manual Section 5.1.5 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.3 Procedures Chapter 3 Requirement: 3.2.3* Procedures. Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established: (1)

• Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program.

BSEP LAR Rev 2 Page A-5

CP&L Attachment A Compliance Statement Compliance Basis Complies with Clarification Procedures are established for inspection, testing and maintenance of fire protection (1) Complies with Clarification systems. Surveillance frequencies are outlined in BSEP plant procedures and may be modified in accordance with the methodology in EPRI Report TR1006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide. Reference Document OPLP-01.2,Fire Protection System Operability, Action, and ALL Surveillance Requirements OAP-033,Fire Protection Program Manual Section 5.3.1.3 Chapter 3 Requirement: 2)

• Compensatory actions implemented when fire protection systems and other systems credited by the fire protection program and this standard cannot perform their intended function and limits on impairment duration.

Compliance Statement Compliance Basis (2) Complies (2) No Additional Clarification. Reference Document DocDetails OPLP-01.2,Fire Protection System Operability, Action, and ALL Surveillance Requirements OAP-033,Fire Protection Program Manual Section 5.3.1.3 OPLP-01.5, Alternative Shutdown Capability Controls ALL Chapter 3 Requirement: (3)

• Reviews of fire protection program - related performance and trends.

Compliance Statement Compliance Basis (3) Complies (3) No Additional Clarification. Reference Document DocDetals EGR-NGGC-0008,Engineering Programs Section 4.6 EGR-NGGC-0010, System & Component Trending Program and Section 1.1 & Attachment 5 System Notebooks Chapter 3 Requirement: (4) Reviews of physical plant modifications and procedure changes for impact on the fire protection program. BSEP LAR Rev 2 Page A-6

CP&L Attachment A Compliance Statement Compliance Basis (4) Complies (4) No Additional Clarification. Reference Document DocDetails OAP-033,Fire Protection Program Manual Section 5.3.4 EGR-NGGC-0003, Design Review Requirements ALL EGR-NGGC-0005,Engineering Change ALL PRO-NGGC-0204, Procedure Review and Approval ALL EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL Chapter 3 Requirement: (5) Long-term maintenance and configuration of the fire protection program. Compliance Statement Compliance Basis (5) Complies (5) No Additional Clarification. Reference Document DoclDtails EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL EGR-NGGC-0003,Design Review Requirements ALL EGR-NGGC-0005,Engineering Change ALL Chapter 3 Requirement: (6) Emergency response procedures for the plant industrial fire brigade. Compliance Statement Compliance Basis (6) Complies (6) No Additional Clarification. Reference Document DoDetails OAP-033,Fire Protection Program Manual Section 3.7, 4.2.17, & 4.2.22 OPFP-013,General Fire Plan Sections 3.3, 3.4, & 3.5 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3 Prevention Chapter 3 Requirement: 3.3 Prevention. A fire prevention program with the goal of preventing a fire from starting shall be established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following: (1) Prevention of fires and fire spread by controls on operational activities. BSEP LAR Rev 2 Page A-7

CP&L Attachment A Compliance Statement Compliance Basis Complies No Additional Clarification (1) Complies (1) No Additional Clarification. Reference Document DoDetai OAP-033,Fire Protection Program Manual Section 5.1.1 OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition ALL Source FIR-NGGC-0003,Hot Work Permit ALL Chapter 3 Requirement: (2) Design controls that restrict the use of combustible materials The design control requirements listed in the remainder of this section shall be provided as described. Compliance Statement Compliance Rank. (2) Complies (2) COMPLIES: No Additional Clarification. Reference Document DocDetails 0FPP-014,Control of Combustible, Transient Fire Loads, and Ignition ALL Source OFPP-013,Transient Fire Load Evaluation ALL 0-89-001 ,Combustible Loading Calculation ALL 2FP-0052,Unit 2 Thermo-Lag Separation Zone Evaluation ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1 Fire Prevention for Operational Activities. Chapter 3 Requirement: 3.3.1 Fire Prevention for Operational Activities. The fire prevention program activities shall consist of the necessary elements to address the control of ignition sources and the use of transient combustible materials during all aspects of plant operations. The fire prevention program shall focus on the human and programmatic elements necessary to prevent fires from starting or, should a fire start, to keep the fire as small as possible. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details 0AP-033,Fire Protection Program Manual Section 5.3 & 5.4 0FPP-014,Control of Combustible, Transient Fire Loads, and Ignition ALL Source BSEP LAR Rev 2 Page A-8

CP&L Attachment A 0FPP-013,Transient Fire Load Evaluation ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.1 General Fire Prevention Activities. Chapter 3 Requirement: 3.3.1.1 General Fire Prevention Activities. The fire prevention activities shall include but not be limited to the following program elements: (1) Training on fire safety information for all employees and contractors including, as a minimum, familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms. Compliance Statement Compliance Basis Complies with Clarification Multiple directives and work practices have been developed to address fire prevention. (1) Complies These directives include but are not limited to the programmatic elements provided in NFPA 805 Section 3.3.1.1 (1) through (3). Upon review of the programmatic elements listed below, BNP believes that the NFPA 805 code requirements are satisfied and no additional elements were evaluated. (1) No Additional Clarification Reference Document DDetails GET SSG,Plant Access and Radiation Worker Training Self Study Section VI & IX Guide GNB01N,Plant Access Initial CBT ALL GNI008N,General Employee Training - Contractors ALL FAQ 06-0028,Training Definition and Content ALL Chapter 3 Requirement: (2)

• Documented plant inspections including provisions for corrective actions for conditions where unanalyzed fire hazards are identified.

Compliance Statement ComplAdiance Basis (2) Complies (2) No Additional Clarification. Reference Document DocDetals~ 0FPP-013,Transient Fire Load Evaluation Section 1 &5.6 HUM-NGGC-0002,Observation Program ALL 0FPP-013,Transient Fire Load Evaluation ALL 0FPP-005,Fire Watch Program ALL BSEP LAR Rev 2 Page A-9

CP&L Attachment A Chapter 3 Requirement: (3)

• Administrative controls addressing the review of plant modifications and maintenance to ensure that both fire hazards and the impact on plant fire protection systems and features are minimized.

Compliance Statement Compliance Basis (3) Complies (3) No Additional Clarification. Reference Document Doc Details EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL OAP-033,Fire Protection Program Manual Section 4.2.8.1 & 5.2.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.2 Control of Combustible Materials Chapter 3 Requirement: 3.3.1.2* Control of Combustible Materials. Procedures for the control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements: (1)

• Wood used within the power block shall be listed pressure-impregnated or coated with a listed fire-retardant application.

Exception: Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 cm) or larger shall not be required to be fire-retardant treated. Compliance Statement Compliance Basis Complies No Additional Clarification (1) Complies (1) No Addition Clarification Reference Document Dotapi OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.3.5.1.b, 6.3.5.1.c, 6.3.5.1.d, 6.3.6 Source 0Al-i14,Housekeeping Program ALL Chapter 3 Requirement: (2) Plastic sheeting materials used in the power block shall be fire-retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale tests, or equivalent. Compliance Statement Compliance Basis (2) Complies (2) No Additional Clarification Reference Document Doc Details BSEP LAR Rev 2 Page A-1 0

CP&L Attachment A FIR-NGGC-0009,NFPA 805 Transient Combustibles and ignition Sections 9.1.9 Source Controls Program Chapter 3 Requirement: (3) Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area immediately following the completion of work or at the end of the shift, whichever comes first. Compliance Statement Compliance Basis (3) Complies (3) No Additional Clarification. Reference Document Doc Details OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.3.5.1 Source OFPP-013,Transient Fire Load Evaluation Section 6.3.1 AIl- 114,Housekeeping Program Sections 6.1.9 & 6.1.10 Chapter 3 Requirement: (4)

• Combustible storage or staging areas shall be designated, and limits shall be established on the types and quantities of stored materials.

Compliance Statement Compliance Basis (4) Complies (4) No Additional Clarification Reference Document Doc Details OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition ALL Source 0FPP-013,Transient Fire Load Evaluation ALL 0-89-001 ,Combustible Loading Calculation ALL Chapter 3 Requirement: (5)

• Controls on use and storage of flammable and combustible liquids shall be in accordance with NFPA 30, Flammable and Combustible Liquids Code, or other applicable NFPA standards.

Compliance Statement Compliance Basis (5) Complies with Clarification (5) COMPLIES WITH CLARIFICATION: FIR-NGGC-0009 establishes controls of Complies via EEEE general housekeeping practices and the control of transient combustibles in the power block. FIR-NGGC-0009 uses NFPA 30 as a developmental reference. FAQ 06-0020 states, in part "This FAQ asks to identify, where used in NFPA 805, Chapter 3, "applicable NFPA standards" for review of programs structures, systems, and components as may be required for Chapter 3 transition using BSEP LAR Rev 2 Page A-1 1

CP&L Attachment A NFPA 805. Because existing fire protection programs for facilities generally provide a listing of NFPA standards used in the development, implementation and maintenance of the fire protection program, the term, "applicable NFPA Standards", where used in NFPA 805, Chapter 3, shall be considered to be equivalent to those NFPA standards identified in the Current License Bases (CLB) for the facility (generally found in the FSAR or approved Fire Protection Program). Because these NFPA standards have been previously approved by the staff for a given facility, this further establishes their applicability. BSEP cites NFPA 30 which applies to the use and storage of flammable and combustible liquids. No other codes which apply to the use and storage of flammable and combustible liquids are cited in the FPPM or CLB. This is therefore acceptable per the guidance in FAQ 06-0020. COMPLIES VIA EEEE: BSEP complies with NFPA 30 as evaluated in code compliance evaluation OFP-0086. This calculation establishes a point-by-point evaluation with NFPA 30. Reference Document DocDetail FIR-NGGC-0009,NFPA 805 Transient Combustibles and Ignition ALL Source Controls Program OFP-0086,Code Compliance Evaluation NFPA 30, Flammable and ALL Combustible Liquids Code FAQ 06-0020,Identification of "applicable NFPA standards" ALL Chapter 3 Requirement: (6)

• Controls on use and storage of flammable gases shall be in accordance with applicable NFPA standards.

Compliance Statement Compliance Basis (6) Complies with Clarification (6) COMPLIES WITH CLARIFICATION: BSEP is not committed to any flammable Complies via EEEE gas standards, and as such are not part of the current license basis. BSEP LAR Rev 2 Page A-1 2

CP&L Attachment A FAQ 06-0020 states, in part "This FAQ asks to identify, where used in NFPA 805, Chapter 3, "applicable NFPA standards" for review of programs structures, systems, and components as may be required for Chapter 3 transition using NFPA 805. Because existing fire protection programs for facilities generally provide a listing of NFPA standards used in the development, implementation and maintenance of the fire protection program, the term, "applicable NFPA Standards", where used in NFPA 805, Chapter 3, shall be considered to be equivalent to those NFPA standards identified in the Current License Bases (CLB) for the facility (generally found in the FSAR or approved Fire Protection Program). Because these NFPA standards have been previously approved by the staff for a given facility, this further establishes their applicability." OFFP-014 establishes controls on the use and storage of flammable gas. OFFP-014 uses NFPA 325M, Fire Hazard Properties of Flammable Liquids, Gases, and Volatile Solids as a developmental reference. COMPLIES VIA EEEE: BSEP complies with controls on use and storage of flammable gases in accordance with NFPA 50A as evaluated in OFP-0090 Reference Document Doc Detail OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.2 Source FAQ 06-0020,Identification of "applicable NFPA standards" ALL OFP-0090,Code Compliance Evaluation NFPA 50A, Standard for ALL Gaseous Hydrogen Systems at Consumer Sites - 1984 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3 Control of Ignition Sources Chapter 3 Requirement: 3.3.1.3 Control of Ignition Sources BSEP LAR Rev 2 Page A-1 3

CP&L Attachment A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.1 [Control of Ignition Sources Code Requirements] Chapter 3 Requirement: 3.3.1.3.1* A hot work safety procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations. Compliance Statement Compliance Basis Complies via EEEE COMPLIES VIA EEEE: Per 0FP-1051, the hot work control processes conform to the Complies with Clarification majority of the applicable requirements of NFPA 51B - 1976, as required by Commitment FH-001. All deviations were reviewed and found to be acceptable. COMPLIES WITH CLARIFICATION: Compliance with NFPA 241 is by clarification and is addressed through compliance with NFPA 51B. NFPA 241, 2009 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot work, states "Responsibility for hot work operations and fire prevention precautions, including permits and fire watches, shall be in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work." Reference Document Doe Detalsa OAP-033,Fire Protection Program Manual 5.3.2 FIR-NGGC-0003,Hot Work Permit ALL OFP-1051,Code Compliance Evaluation NFPA 51B, Standard for Fire ALL Prevention in Use of Cutting and Welding Processes NFPA 241,Standard for Safeguarding Construction, Alteration, and Section 5.1.1 Demolition Operations, 2009 Edition NED-M/BMRK-0001,NFPA 51B Code Compliance Evaluation for ALL BSEP LAR Rev 2 Page A-14

CP&L Attachment A NFPA 51 B, Standard for Fire Prevention during Welding, Cutting, and Other Hot Work - 1999 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.2 [Control of Ignition Sources on Smoking Limitations] Chapter 3 Requirement: 3.3.1.3.2 Smoking and other possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document DoDetals OAP-033,Fire Protection Program Manual Section 5.3.2.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.3 [Control of Ignition Sources for Leak Testing] Chapter 3 Requirement: 3.3.1.3.3 Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document DoDetails OAP-033,Fire Protection Program Manual Section 5.3.2.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.4 [Control of Ignition Sources on Portable Heaters] Chapter 3 Requirement: 3.3.1.3.4* Plant administrative procedure shall control the use of portable electrical heaters in the plant. Portable fuel-fired heaters shall not be permitted in plant areas containing equipment important to nuclear safety or where there is a potential for radiological releases resulting from a fire. BSEP LAR Rev 2 Page A-1 5

CP&L Attachment A Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document noc Mails OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.5 Source FIR-NGGC-0009,NFPA 805 Transient Combustibles and Ignition Section 9.1.11 Source Controls Program Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.2 Structural. Chapter 3 Requirement: 3.3.2 Structural. Walls, floors, and components required to maintain structural integrity shall be of noncombustible construction, as defined in NFPA 220, Standard on Types of Building Construction. Compliance Statement Compliance Basis Complies COMPLIES: No Additional Clarification. Complies via EEEE The referenced evaluation 0FP-0033, evaluated the acceptability of not fire proofing exposed structural steel which is located in the control building elevator shaft. This evaluation is limited to the following fire zones; CB-6, CB-1 1 and CB-22 which are located within fire area CB-23E. The evaluation concluded that the steel columns installed in the west wall of the Control Building elevator shaft have adequate fire resistance for the worst case fires expected in either the elevator shaft or the men's restroom if left unprotected. As such the "complies via EEEE" compliance strategy applies only to Fire Area CB-23E. All other Fire Areas fall under the "complies" compliance strategy. Reference Document NFPA 220,Standard on Types of Building Construction, Online Section 3.3.4 Edition 2009 APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.1, page 5 Plant, January 1, 1977 0FP-0033,Structural Steel Fireproofing ALL Page A-16 Rev 22 BSEP LAR Rev Page A-1 6

Attachment A TableB1NP80Ch3TastoDeal Table B-i NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.3 Interior Finishes Chapter 3 Requirement: 3.3.3 Interior Finishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101, Life Safety Code, requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class I interior floor finishes. Compliance Statement Compliance Basis Complies via EEEE COMPLIES VIA EEEE: BSEP complies with NFPA 101 Code Complies with Clarification Requirements for interior finishes as evaluated in EER 94-0009 & ESR 99-Complies 00109. COMPLIES WITH CLARIFICATION - Per the BSEP UFSAR, the following fire zones have carpeting listed as fixed combustibles: CB- 16 (Office Area), CB- 19 (Central Alarm Station), CB-20 (Northwest Back Panel Zone), CB-21 (Southwest Back Panel Zone), RW-1 B (Radwaste CFD Area). While specific documentation does not exist for CB-16, CB-19, and RW-1B, the requirements and standards associated with their application have remained consistent under NEIL insurance such that reasonable assurance is present that the carpet selected would meet NFPA 253 Class I requirements. Configuration controls in place would continue to ensure that this is consistent for any future installations or replacements. Reference Document Doc Details NFPA 101,Life Safety Code, 2003 Edition Sections 10.2.3.4(1), 10.2.7.3, 10.2.7.4 APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.1-5 Plant, January 1, 1977 CPL-XXXX-W-005,Nuclear Power Plant Protective Coatings Appendix A UFSAR,Updated Final Safety Analysis Report Section 9.5.1.5 EER 94-0009,Evaluation of Floor Coatings on Combustible Loading ALL AMERCOATAmerlock 400 Series Data Sheet Qualifications Section BSEP LAR Rev 2 Page A-1 7

CP&L Attachment A ESR 99-00109,Control Room Carpet Additions ALL EC 47763,Control Room Project ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.4 Insulation Materials Chapter 3 Requirement: 3.3.4 Insulation Materials. Thermal insulation materials, radiation shielding materials, ventilation duct materials, and soundproofing materials shall be noncombustible or limited combustible. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Documentc Detils APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.1, page 5 Plant, January 1, 1977 Specification 226-001 ,Sheet Metal Work and Accessories Section III Specification 226-002,Sheet Metal Work Section III Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5 Electrical. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5.1 [Electrical Wiring Above Suspended Ceiling Limitations] Chapter 3 Requirement: 3.3.5.1 Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be BSEP LAR Rev 2 Page A-1 8

CP&L Attachment A listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers. Compliance Statement Compliance Basis Complies The Fire Protection Program Review submitted under Docket No. 50-325 & 50-324 on January 1, 1977 concluded that suspended ceilings and their supports are of non-combustible construction. Concealed spaces are devoid of combustibles. Reference Document Doc Details APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.1 - 7 Plant, January 1, 1977 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5.2 [Electrical Raceway Construction Limits] Chapter 3 Requirement: 3.3.5.2 Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components. Compliance Statement Compliance Basis Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC APPROVAL: Runs of flexible metallic Complies with Clarification conduits exist within the cable accessways. These runs were reviewed by the NRC in a 1977 Safety Evaluation Report. Their review states:

                                                 "5.1.2 Combustibles About 20 percent of the conduit is flexible greenfield type with a polyvinyl-chloride (PVC) covering. This is the only exposed combustible material located here.

5.1.6 The licensee has proposed to: (1) Install a fire wall having a three-hour rating between the redundant division conduits to assure that a single fire will not involve redundant safe shutdown systems. (2) Coat the polyvinyl-chloride covered conduit with a flame retardant coating to BSEP LAR Rev 2 Page A-1 9

CP&L Attachment A limit the consequences of a fire. (3) Provide water hose stations within easy access to these areas for additional suppression capability. (4) Provide additional detector for the new areas created by the addition of walls so that at least two detectors are in each fire area. We conclude that, subject to implementation of the above described modifications, fire protection for the cable accessways satisfies the objectives identified in Section 2.1 of this report and is, therefore, acceptable." Per the Updated Final Safety Analysis Report (UFSAR), Rev. 23, fire walls having three-hour ratings separate redundant division conduits, flame-retardant coatings have been applied to conduit and cable trays in cable access ways and spreading areas, each access way has a water hose station within easy access to the area for additional suppression capability, and each access way has fire detectors in the zone to provide prompt notification of a fire. COMPLIES WITH CLARIFICATION: Cable drops as described in Specification 048-001 align with the guidance of FAQ 0021. The compliance strategy "Complies via Previous NRC Approval" applies only to the following Control Building cable accessway fire zones: CB-01A, CB-01B, CB-02A, CB-02B, CB-12A, CB-12B, CB-13A and CB- 13B for existing electrical raceway construction details which are located in Fire Areas CB-1 and CB-2. The "Complies with Clarification" compliance statement applies to all other plant fire areas and zones relative to Specification 048-001 requirements and the guidance of FAQ 06-0021. As such the "Complies via Previous NRC Approval" compliance strategy applies only to Fire Areas CB-1 and CB-2. All other fire areas fall under the "Complies with Clarification" compliance strategy. BSEP LAR Rev 2 Page A-20

CP&L Attachment A Reference Document Doc Details Fire Protection Safety Evaluation Report,Fire Protection Safety Sections 5.1.2 & 5.1.6 Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 UFSAR,Updated Final Safety Analysis Report Sections 9.5.1.5 & 9.5.1.4.3.4.2.b FAQ 06-0021,Cable Air Drops ALL Specification 048-001 Installation of the Electrical Raceway System Sections 2.2.1.2, 2.2.2, 2.2.11.2, 2.2.12.10 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5.3 [Electrical Cable Flame Propagation Limits] Chapter 3 Requirement: 3.3.5.3* Electric cable construction shall comply with a flame propagation test as acceptable to the AHJ. Compliance Statement Compliance Bases Complies via Previous NRC Approval COMPLIES WITH CLARIFICATION: FAQ 06-0022 provides an Appendix to evaluate Complies with Clarification currently recognized flame propagation tests to the IEEE 383-1974 Standard, the US NRC minimum test standard, and acceptance criteria for cable flame propagation tests. The specifications applicable for the procurement of various cables are listed in Section 3.2.5 of DBD-112. Per DBD-112, cables procured during and following construction were qualified as being self-extinguishing and non-propagating and they meet or exceed the IEEE 383 flame test. COMPLIES VIA PREVIOUS NRC APPROVAL: Per the SER dated 11/22/77, the NRC had the following finding:

                                                  "Flame tests conducted on the electrical cables used in the Brunswick Plant were comparable to the combustibility test set forth in IEEE 383. The results show that in the configurations and with the ignition source used in the tests the cable insulation burns slowly. Nevertheless, we consider all cable insulation made of organic material as combustible and, therefore, we find that the retest to the IEEE 383 procedures and criteria would not provide information that would alter our BSEP LAR Rev 2                                                                                                               Page A-21

CP&L Attachment A conclusions. Accordingly, we find the electrical cables used at the Brunswick Plant acceptable." Reference Document DBD-112,Cables and Raceways Sections 2.1.3.3 &3.2.5 FAQ 06-0022,Acceptable Electrical Cable Construction Tests ALL Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.8 Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 NED-B/BOP-1001,Evaluateds UL-910 Qualified Cables as Substitute ALL for IEEE-383 Rated Cables; Generic for all Plants 86-0378,Evaluation of Deleting Requirement for Fire Retardant ALL Coating of Telephone Cable 90-0334,Acceptance Criteria for Cable Coatings; Supersedes EER ALL 89-0056, Rev. 1 Table B-i NFPA 805 Ch.3 Transition Details Chapter 3 References 3.3.6 Roofs. Chapter 3 Requirement: 3.3.6 Roofs. Metal roof deck construction shall be designed and installed so the roofing system will not sustain a self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building. Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods of Fire Tests of Roof Coverings. Compliance Statement Compliance Basis Complies with Clarification NFPA 256 was not an original design requirement for the plant, was not referenced in BTP 9.5-1, and was not a condition in previous NRC Safety Evaluation Reports. Therefore, BSEP was never evaluated to NFPA 256 but met the equivalent requirement in BTP 9.5-1, which is that metal deck roof construction should be non-combustible (see the building materials directory of the Underwriters' Laboratory, Inc.) or listed as Class I by Factory Mutual System Approval Guide (A Factory Mutual Class I roof is considered equivalent to NFPA 256 Class A roof classification). In its letter to NRC dated 12-29-1976, BSEP stated "Metal deck roof construction is non-combustible and complies with the BSEP LAR Rev 2 Page A-22

CP&L Attachment A requirements of Class I construction of Factory Mutual Standards." Reference Document Doc Details APCSB 9.5-1 ,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3d.1-6 Plant, January 1, 1977 Table B-i NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7 Bulk Flammable Gas Storage. Chapter 3 Requirement: 3.3.7 Bulk Flammable Gas Storage. Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.2, page 4 Plant, January 1, 1977 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.1 [Bulk Flammable Gas Location Requirements] Chapter 3 Requirement: 3.3.7.1 Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire or explosion will not adversely impact systems, equipment, or components important to nuclear safety. NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed for hydrogen storage. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.2, page 3-4 Plant, January 1, 1977 OFP-0090,Code Compliance Evaluation NFPA 50A, Standard for ALL Gaseous Hydrogen Systems at Consumer Sites - 1984 Edition BSEP LAR Rev 2 Page A-23

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.2 [Bulk Flammable Gas Container Restrictions] Chapter 3 Requirement: 3.3.7.2 Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is not pointed at buildings. Compliance Statement Compmiance Basis Complies No Additional Clarification Reference Document DoDetails APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.d.2, page 4 Plant, January 1, 1977 OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.2.1.3 Source Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.3 [Bulk Flammable Gas Cylinder Limitations] Chapter 3 Requirement: 3.3.7.3 Flammable gas storage cylinders not required for normal operation shall be isolated from the system. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.2 Source Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.8 Bulk Storage of Flammable and Combustible Liquids. Chapter 3 Requirement: 3.3.8 Bulk Storage of Flammable and Combustible Liquids. Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing systems, equipment, or components important to nuclear safety. As a minimum, storage and use shall BSEP LAR Rev 2 Page A-24

CP&L Attachment A comply with NFPA 30, Flammable and Combustible Liquids Code. Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 30 as evaluated in 0FP-0086. Reference Document DoDafls OFP-0086,Code Compliance Evaluation NFPA 30, Flammable and ALL Combustible Liquids Code Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.9 Transformers. Chapter 3 Requirement: 3.3.9* Transformers. Where provided, transformer oil collection basins and drain paths shall be periodically inspected to ensure that they are free of debris and capable of performing their design function. Compliance Statement Compliance Basis Complies with Clarification See Implementation Item pertinent to NFPA 805 Chapter 3, Section 3.3.9 compliance in Attachment S of the Transition Report. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.10 Hot Pipes and Surfaces. Chapter 3 Requirement: 3.3.10* Hot Pipes and Surfaces. Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact with hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document DoDeti OFPP-014,Control of Combustible, Transient Fire Loads, and Ignition Section 6.3.1.1 Source BSEP LAR Rev 2 Page A-25

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3 References 3.3.11 Electrical Equipment Chapter 3 Requirement: 3.3.11 Electrical Equipment Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document DoclDtails FIR-NGGC-0009,NFPA 805 Transient Combustibles and Ignition Section 9.1.12 Source Controls Program Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.12 Reactor Coolant Pumps. Chapter 3 Requirement: 3.3.12* Reactor Coolant Pumps. For facilities with non-inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply. (1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil system. Compliance Statement Compliance Basis (1) N/A (1) N/A Chapter 3 Requirement: (2) Leakage shall be collected and drained to a vented closed container that can hold the inventory of the reactor coolant pump lubricating oil system. Compliance Statement Compliance Basis (2) N/A (2) N/A Chapter 3 Requirement: (3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazard of a fire flashback. Compliance Statement Compliance Basis BSEP LAR Rev 2 Page A-26

CP&L Attachment A (3) N/A (3) N/A Chapter 3 Requirement: (4) Leakage points on a reactor coolant pump motor to be protected shall include but not be limited to the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps. Compliance Statement Compliance Basis (4) N/A (4) N/A Chapter 3 Requirement: (5) The collection basin drain line to the collection tank shall be large enough to accommodate the largest potential oil leak such that oil leakage does not overflow the basin. Compliance Statement Compliance Basis (5) N/A (5) N/A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4 Industrial Fire Brigade. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.1 On-Site Fire-Fighting Capability. Chapter 3 Requirement: 3.4.1 On-Site Fire-Fighting Capability. All of the following requirements shall apply. (a) A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable: (1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting) Compliance Statement Compliance Basis (a) Complies (a) No Additional Clarification BSEP LAR Rev 2 Page A-27

CP&L Attachment A (a)(1) Complies via EEEE (a)(1): BSEP complies with NFPA 600 as evaluated in NED-M/BMRK-0002. Reference Document Doc Details OPLP-01.2,Fire Protection System Operability, Action, and Section 6.11.1 Surveillance Requirements OFPP-008,Fire Protection Equipment Monthly Inspection Sections 1 & 3.1 FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program ALL NED-M/BMRK-0002,Code Compliance Evaluation for NFPA 600, ALL Standard on Industrial Fire Brigades, 2000 Edition Chapter 3 Requirement: (2) NFPA 1500, Standard on Fire Department Occupational Safety and Health Program Compliance Statement Compliance Basis (a)(2) N/A (a)(2): NFPA 1500 is not applicable to BSEP per FAQ 06-0007 which states," The NFPA standards divide fire brigades into two types, based on organization and duties: "Industrial Fire Brigades" and "Industrial Fire Departments." Practically, this means that a fire fighting organization at a nuclear power plant must comply with either NFPA 600 (for an Industrial Fire Brigade) or both NFPA 1500 and NFPA 1582 (for an Industrial Fire Department)." BSEP will show compliance with NFPA 600. Reference Document DoDetails FAQ 06-0007,NFPA -805 Section 3.4.1, Specific Clarification ALL Chapter 3 Requirement: (3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for Fire Department Physicians. Compliance Statement Compliance RBas (a)(3) N/A (a)(3): NFPA 1582 is not applicable to BSEP per FAQ 06-0007 which states," The NFPA standards divide fire brigades into two types, based on organization and duties: "Industrial Fire Brigades" and "Industrial Fire Departments." Practically, this means that a fire fighting organization at a nuclear power plant must comply with either NFPA 600 (for an Industrial Fire Brigade) or both NFPA 1500 and NFPA BSEP LAR Rev 2 Page A-28

CP&L Attachment A 1582 (for an Industrial Fire Department)." BSEP will show compliance with NFPA 600. Reference Document D FAQ 06-0007,NFPA -805 Section 3.4.1, Specific Clarification ALL Chapter 3 Requirement: (b)

• Industrial fire brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required.

Compliance Statement Compliance Basis (b) Complies (b) No Additional Clarification Reference Document DoDetals OAP-033,Fire Protection Program Manual Section 5.5.2.c OPLP-01.2,Fire Protection System Operability, Action, and Section 6.11.1 Surveillance Requirements Chapter 3 Requirement: (c) During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance Exception: Sufficient training and knowledge shall be permitted to be provided by an operations advisor dedicated to industrial fire brigade support criteria. Compliance Statement Compliance Basis (c) Complies (c) No Additional Clarification Reference Document DoDetails OAP-033,Fire Protection Program Manual Sections 4.2.22, 4.3.9, 5.5.1.2.b OFPP-031,Fire Brigade Staffing Roster and Equipment Requirements Section 3.1.1 FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program all Chapter 3 Requirement: (d)

• The industrial fire brigade shall be notified immediately upon verification of a fire.

Compliance Statement Compliance Basis (d) Complies (d) No Additional Clarification Reference Document DoDails OPFP-013,General Fire Plan Section 3.2.3 Chapter 3 Requirement: (e) Each industrial fire brigade member shall pass an annual physical examination to determine that he BSEP LAR Rev 2 Page A-29

CP&L Attachment A or she can perform the strenuous activity required during manual fire-fighting operations. The physical examination shall determine the ability of each member to use respiratory protection equipment. Compliance Statement Compliance Basis (e) Complies (e) No Additional Clarification Reference Document Doc Details OAP-033,Fire Protection Program Manual Section 4.3.3 29CFR1910.156,Labor; Occupational Safety and Health Section b(2) Administration, Department of Labor; Occupational Safety and Health Standards; Fire Brigades, Rev. as of 7/1/2002 OPLP-01.2,Fire Protection System Operability, Action, and Section 6.11.4.3 Surveillance Requirements FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.9.2 &9.4.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2 Pre-Fire Plans. Chapter 3 Requirement: 3.4.2* Pre-Fire Plans. Current and detailed pre-fire plans shall be available to the industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document OAP-033,Fire Protection Program Manual Sections 4.2.16.3, 5.6.3 OFPP-008,Fire Protection Equipment Monthly Inspection Attachment 2 OPFP-CB,Control Building Prefire Plans ALL OPFP-DG,Diesel Generator Building Prefire Plans ALL OPFP-MBOCA,Miscellaneous Buildings - Owner Controlled Area ALL OPFP-MBPA,Miscellaneous Buildings Prefire Plans - Protected Area ALL OPFP-PBAA,Power Block Auxiliary Areas Prefire Plans (SW, RW, ALL AOG, T, EY, PDC) 1PFP-RB,Reactor Building Prefire Plans ALL 1PFP-TB,Turbine Building Prefire Plans ALL 2PFP-RB,Reactor Building Prefire Plans ALL 2PFP-TB,Turbine Building Prefire Plans ALL FIR-NGGC-0008,NFPA 805 Pre-Fire Plans ALL BSEP LAR Rev 2 Page A-30

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2.1 [Pre-Fire Plan Contents] Chapter 3 Requirement: 3.4.2.1* The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area, along with any nuclear safety components and fire protection systems and features that are present. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Detaiia OAP-033,Fire Protection Program Manual Section 5.6.3.2 OPFP-013,General Fire Plan ALL FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 9.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2.2 [Pre-Fire Plan Updates] Chapter 3 Requirement: 3.4.2.2 Pre-fire plans shall be reviewed and updated as necessary. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DoDe 0AP-033,Fire Protection Program Manual Sections 4.2.16.3, 4.2.8.1, & 5.3.3.1 PRO-NGGC-0204,Procedure Review and Approval ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2.3 [Pre-Fire Plan Locations] Chapter 3 Requirement: 3.4.2.3* Pre-fire plans shall be available in the control room and made available to the plant industrial fire brigade. Compliance Statement Compliance Basis BSEP LAR Rev 2 Page A-31

CP&L Attachment A Complies No Additional Clarification. Reference Document Dochmetalsa OFPP-008,Fire Protection Equipment Monthly Inspection Attachment 2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3 Reference* 3.4.2.4 [Pre-Fire Plan Coordination Needs] Chapter 3 Requirement: 3.4.2.4* Pre-fire plans shall address coordination with other plant groups during fire emergencies. Compliance Statement Compliance Basis Complies with Clarification Coordination with other plant groups during fire emergencies is described in 0PFP-013. Site procedure OPFP-013, is not specifically a fire pre-plan however OPFP-013 provides specific instructions for actions required from key groups at BSEP supporting the fire brigade/fire emergency actions. There are detailed response coordination actions specified for Control Room personnel, Security group, and Health Physics group. Any other coordination actions would be initiated by Control Room personnel as needed for any plant emergency. Reference Document DoDetail OPFP-013,General Fire Plan ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.3 Training and Drills. Chapter 3 Requirement: 3.4.3 Training and Drills. Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities. (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply. (1) Plant industrial fire brigade members shall receive training consistent with the requirements BSEP LAR Rev 2 Page A-32

CP&L Attachment A contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate. Compliance Statement Compliance Basis (a)(1) Complies via EEEE (a)(1): BSEP complies with NFPA 600 as evaluated in NED-M/BMRK-0002. Reference Document DoDetals NED-M/BMRK-0002,Code Compliance Evaluation for NFPA 600, ALL Standard on Industrial Fire Brigades, 2000 Edition Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(2) Industrial fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire. Compliance Statement Compliance Basis (a)(2) Complies (a)(2): No Additional Clarification Reference Document Doc etall GNR01N,Plant Access Annual Requalification, CBT ALL GNR02N,Rad Worker Annual Requalification, CBT ALL FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Sections 9.8 & 9.10 Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(3) A written program shall detail the industrial fire brigade training program. Compliance Statement Compliance Basis (a)(3) Complies (a)(3) No Additional Clarification Reference Document DoDetals FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program ALL Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(4) Written records that include but are not limited to initial industrial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill attendance records, and leadership training for industrial fire brigades shall be maintained for each industrial fire brigade member. Compliance Statement Compliance Basis (a)(4) Complies (a)(4) No Additional Clarification Reference Document DocDeaals~ BSEP LAR Rev 2 Page A-33

CP&L Attachment A TAP-404,Training Documentation and Records ALL TAP-416,Fire Protection Training Administrative Procedure ALL FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program ALL Chapter 3 Requirement: (b) Training for Non-Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial fire brigade shall be trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade. Compliance Statement Compliance Basis (b) Complies with Clarification (b) Guidance for non-industrial fire brigade members is found in OPFP-013. The procedure defines the actions needed to be taken by personnel discovering a fire, security personnel actions, duty health physics contact actions, and duty maintenance contact actions. Reference Document DoDetals OPFP-013,General Fire Plan ALL FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program ALL Chapter 3 Requirement: (c)

• Drills. All of the following requirements shall apply.

(1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire brigade. Compliance Statement Compliance Basis (c)(1) Complies (c)(1) No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.10.3.3.a Chapter 3 Requirement: (c)

• Drills. All of the following requirements shall apply.(2) Industrial fire brigade drills shall be developed to test and challenge industrial fire brigade response, including brigade performance as a team, proper use of equipment, effective use of pre-fire plans, and coordination with other groups.

These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario. Compliance Statement Compliance Basis (c)(2) Complies (c)(2) No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.10 BSEP LAR Rev 2 Page A-34

CP&L Attachment A Chapter 3 Requirement: (c)

• Drills. All of the following requirements shall apply.(3) Industrial fire brigade drills shall be conducted in various plant areas, especially in those areas identified to be essential to plant operation and to contain significant fire hazards.

Compliance Statement Compliance Basis (c)(3) Complies (c)(3) No Additional Clarification Reference Document DoDetals FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Sections 9.10.2 Chapter 3 Requirement: (c)

• Drills. All of the following requirements shall apply.(4) Drill records shall be maintained detailing the drill scenario, industrial fire brigade member response, and ability of the industrial fire brigade to perform as a team.

Compliance Statement Compliance Basis (c)(4) Complies (c)(4) No Additional Clarification Reference Document Doc Details TAP-404,Training Documentation and Records ALL FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program ALL Chapter 3 Requirement: (c)

• Drills. All of the following requirements shall apply.(5) A critique shall be held and documented after each drill.

Compliance Statement Compliance Basis (c)(5) Complies (c)(5) No Additional Clarification Reference Document Doc Details TAP-404,Training Documentation and Records ALL FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.10.7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.4 Fire-Fighting Equipment. Chapter 3 Requirement: 3.4.4 Fire-Fighting Equipment. Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards. BSEP LAR Rev 2 Page A-35

CP&L Attachment A Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 600 Fire Brigade Equipment requirements as evaluated in NED-M/BMRK-0002. Reference Document OFPP-039,SCBA Use and Maintenance ALL OFPP-031,Fire Brigade Staffing Roster and Equipment Requirements ALL OFPP-008,Fire Protection Equipment Monthly Inspection ALL NED- M/BMRK-0002,Code Compliance Evaluation for NFPA 600, Code Sections 2-6, 2-7, 4-3, 5-3 Standard on Industrial Fire Brigades, 2000 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5 Off-Site Fire Department Interface. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.1 Mutual Aid Agreement. Chapter 3 Requirement: 3.4.5.1 Mutual Aid Agreement. Off-site fire authorities shall be offered a plan for their interface during fires and related emergencies on site. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DocDe OERP,Radiological Emergency Response Plan Sections 3.2.1, 3.7.2, & Appendix B FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.7 BSEP LAR Rev 2 Page A-36

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.2 Site-Specific Training. Chapter 3 Requirement: 3.4.5.2* Site-Specific Training. Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited to participate in a drill at least annually. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document OPEP-04.3,Performance of Training, Exercises, and Drills Section 6.7.1 FIR-NGGC-0007,NFPA 805 Fire Brigade Training Program Section 9.7.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.3 Security and Radiation Protection. Chapter 3 Requirement: 3.4.5.3* Security and Radiation Protection. Plant security and radiation protection plans shall address off-site fire authority response. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document DoDataia OPFP-013,General Fire Plan Sections 3.6 & 3.7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.6 Communications. Chapter 3 Requirement: 3.4.6* Communications. An effective emergency communications capability shall be provided for the industrial fire brigade. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document nor ntamis BSEP LAR Rev 2 Page A-37

CP&L Attachment A 0AP-033,Fire Protection Program Manual Section 5.5.4.3 001-01.02,Operations Unit Organization and Operating Practices Section 5.5.5. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5 Water Supply Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.1 [Water Supply Flow Code Requirements] Chapter 3 Requirement: 3.5.1 A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of the two following methods. (a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies. (b) Calculate the fire flow rate for 2 hours. This fire flow rate shall be based on 500 gpm (1892.5 L/min) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system(s) in the power block as determined in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service. Compliance Statement Compliance Basis Complies via Previous NRC Approval The licensing basis for the water storage tanks is that contained in Appendix A to BTP APCSB 9.5-1. This guideline stated:

                                               "Two separate, reliable water supplies should be provided. If tanks are used, two 100% (minimum of 300,000 gallons each) system capacity tanks should be installed.

They should be so interconnected that pumps can take suction from either or both. However, a leak in one tank or its piping should not cause both tanks to drain. The main plant fire water supply BSEP LAR Rev 2 Page A-38

CP&L Attachment A capacity should be capable of refilling either tank in a minimum of eight hours. Common tanks are permitted for fire and sanitary or service water storage. When this is done, however, minimum fire water storage requirements should be dedicated by means of a vertical standpipe for other water services." The 300,000 and 200,000 gallon water tanks, as approved by the SE Report, are still used as the source of fire water at BSEP as described in the SE Report. There have been no plant modifications or other changes that would invalidate the basis for approval. The fire protection water supply system has not been changed which would affect the capacity to provide the required supply. Reference Document DocDetails OPLP-01 .2,Fire Protection System Operability, Action, and Section 6.1.1 & 6.1.3.f Surveillance Requirements DBD-62,Water Based Suppression System Section 3.3.2 OPT-34.7.1.0, Fire Suppression Water System Flow Test ALL Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1 (1) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 Table B-i NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.2 [Water Supply Tank Code Requirements] Chapter 3 Requirement: 3.5.2* The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection. Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction from a large body of water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated. Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service. Compliance Statement Compliance Basis Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC BSEP LAR Rev 2 Page A-39

CP&L Attachment A APPROVAL: Per Fire Protection Safety Complies via EEEE Evaluation Report, Brunswick Nuclear Plant, November 22, 1977, "The fire protection water supply for both reactor units consists of two fire pumps taking suction from a single 300,000 gallon storage tank. As a backup supply, the pumps can also take suction from a 200,000 gallon demineralized water tank by manually operating a normally closed gate valve at the pumps. The demineralized water tank is not reserved for fire protection but could be made available manually. The primary fire water tank is large enough to provide over two hours of fire flow for the largest expected demand and its water level is electronically supervised. In the unlikely event of a catastrophic leak in the primary tank, the secondary supply could provide an adequate supply of water for suppressing fires in safety-related areas. We conclude that the water supply for fire protection satisfies the objectives identified in Section 2.1 of this report and is, therefore, acceptable." The licensing basis for the water storage tanks is that contained in Appendix A to BTP APCSB 9.5-1. This guideline stated:

                 "Two separate, reliable water supplies should be provided. If tanks are used, two 100% (minimum of 300,000 gallons each) system capacity tanks should be installed.

They should be so interconnected that pumps can take suction from either or both. However, a leak in one tank or its piping should not cause both tanks to drain. The main plant fire water supply capacity should be capable of refilling either tank in a minimum of eight hours. Common tanks are permitted for fire and sanitary or service water storage. When this is done, however, minimum fire water storage requirements should be dedicated by means of a vertical standpipe for other water services." The 300,000 and 200,000 gallon water BSEP LAR Rev 2 Page A--40

CP&L Attachment A tanks, as approved by the SE Report, are still used as the source of fire water at BSEP as described in the SE Report. There have been no plant modifications or other changes that would invalidate the basis for approval. The fire protection water supply system has not been changed which would affect the capacity to provide the required supply. BSEP does not utilize Exception 1 or 2 of NFPA 805 Section 3.5.2. COMPLIES VIA EEEE: BSEP complies with NFPA 22 as evaluated in BSEP Calculation 0FP-0089. Reference Document DocDetaols Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1(1) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 DBD-62,Water Based Suppression System Sections 3.3.1 & 3.6.3 D-04106,Plant Fire Protection System Piping Diagram ALL 0FP-0089,Code Compliance Evaluation NFPA 22, Water Tanks for ALL Private Fire Protection - 1971 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.3 [Water Supply Pump Code Requirements] Chapter 3 Requirement: 3.5.3* Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source. Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 20 as evaluated in 0FP-1018. Reference Document DoDetails 0FP-1018,Code Compliance Evaluation NFPA 20 Conclusion Section & Attachment 4 (Code Section 32) Table B-1 NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A-41

CP&L Attachment A Chapter 3

Reference:

3.5.4 [Water Supply Pump Diversity and Redundancy] Chapter 3 Requirement: 3.5.4 At least one diesel engine-driven fire pump or two more seismic Category I Class IE electric motor-driven fire pumps connected to redundant Class IE emergency power buses capable of providing 100 percent of the required flow rate and pressure shall be provided. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document OPLP-01.2,Fire Protection System Operability, Action, and Section 6.1.1 &6.1.3 Surveillance Requirements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.5 [Water Supply Pump Separation Requirements] Chapter 3 Requirement: 3.5.5 Each pump and its driver and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers. Compliance Statement Compliance Basis Complies via Previous NRC Approval Per Fire Protection Safety Evaluation Report, Brunswick Nuclear Plant, November 22, 1977, "Both fire pumps and their controllers are located in the water treatment building, and could be subject to damage by a fire in that structure. To preclude such an event, the licensee has proposed to provide automatic sprinklers, and barriers, to prevent flame impingement between the pumps and between the pumps and the controllers, and three hour fire barriers between the building and the diesels fuel tank. A flow switch and cutoff valve to detect a rupture in the supply line and shut off fuel flow to the diesel driven fire will be provided. We conclude that, subject to the implementation of the above described modifications, the fire pumps satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable. BSEP LAR Rev 2 Page A-42

CP&L Attachment A Reference Document DBD-62,Water Based Suppression System Section 3.4.1 Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1(2), page 4-2 Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 NG-77-706,Fire Protection Program Evaluation Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.6 [Water Supply Pump Start/Stop Requirements] Chapter 3 Requirement: 3.5.6 Fire pumps shall be provided with automatic start and manual stop only. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Specification 238-016,Pumps and Accessories for Fire Protection Section II - E.1 & E.2 Section III - B.3.b.2 & C.3.c System, 3/29/1976 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.7 [Water Supply Pump Connection Requirements] Chapter 3 Requirement: 3.5.7 Individual fire pump connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details D-04106,Plant Fire Protection System Piping Diagram ALL D-02043-SHOO01, Plant Fire Protection System Piping Diagram ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.8 [Water Supply Pressure Maintenance Limitations] BSEP LAR Rev 2 Page A-43

CP&L Attachment A Chapter 3 Requirement: 3.5.8 A method of automatic pressure maintenance of the fire protection water system shall be provided independent of the fire pumps. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DoDetails DBD-62,Water Based Suppression System Section 0.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.9 [Water Supply Pump Operation Notification] Chapter 3 Requirement: 3.5.9 Means shall be provided to immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DocDetail F-07350,UA-35, UA-36, & UA-37 Annunciator Lamp Boxes Window ALL Arrangement Specification 070-011 ,Fire Detection System Attachment 4, Page 5 of 6 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.10 [Water Supply Yard Main Code Requirements] Chapter 3 Requirement: 3.5.10 An underground yard fire main loop, designed and installed in accordance with NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish anticipated water requirements. Compliance Statement Compliance Basis Complies via EEEE BNP complies with NFPA 24 as evaluated in 0FP-1017. Reference Document Doc Details BSEP LAR Rev 2 Page A-44

CP&L Attachment A OFP-1017,Code Compliance Evaluation NFPA 24 ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.11 [Water Supply Yard Main Maintenance Issues] Chapter 3 Requirement: 3.5.11 Means shall be provided to isolate portions of the yard fire main loop for maintenance or repair without simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations provided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems. Compliance Statement Compliance Basis Complies COMPLIES WITH CLARIFICATION: The current design of the MWT Complies with Clarification automatic/manual water based fire suppression system requires the simultaneous shutoff to both the fixed fire suppression system as well as the single fire hose station provided for manual backup for maintenance or repair. Both the primary and backup fire suppression systems would be impaired in the case of a single active failure/crack in the water supply piping. Adequate means to combat a fire by the fire brigade is provided via a hydrant located close to the MWT Building. Reference Document nlaiei DBD-62,Water Based Suppression System Section 3.3.3 &3.6.3 D-02058-SH003B,Plant Fire Protection System Piping Diagram ALL D-02304,Piping Diagram Service Water Radwaste & Treatment ALL Buildings Fire Protection Sprinkler System Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.12 [Water Supply Compatible Thread Connections] Chapter 3 Requirement: 3.5.12 Threads compatible with those used by local fire departments shall be provided on all hydrants, hose BSEP LAR Rev 2 Page A-45

CP&L Attachment A couplings, and standpipe risers. Exception: Fire departments shall be permitted to be provided with adapters that allow interconnection between plant equipment and the fire department equipment if adequate training and procedures are provided. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DDetails APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.b-12 Plant, January 1, 1977 DBD-62,Water Based Suppression System Section 3.3.11 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.13 [Water Supply Header Options] Chapter 3 Requirement: 3.5.13 Headers fed from each end shall be permitted inside buildings to supply both sprinkler and standpipe systems, provided steel piping and fittings meeting the requirements of ANSI B31.1, Code for Power Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems where such headers are part of the seismically analyzed hose standpipe system. Where provided, such headers shall be considered an extension of the yard main system. Each sprinkler and standpipe system shall be equipped with an outside screw and yoke (OS&Y) gate valve or other approved shutoff valve. Compliance Statement Compliance Basis Complies COMPLIES WITH CLARIFICATION: Although no individual isolation valves are Complies with Clarification provided for each standpipe connection supplied by the Turbine Building Fire Protection Headers, each individual fixed sprinkler system can be isolated from the Turbine Building Fire Protection Header (for maintenance/repair) without simultaneously impairing manual fire suppression capabilities. Reference Document Doc Deta~il Specification 248-117,Specification for Installation of Piping Systems Section 2.10 D-02057 Sh. 2A,Plant Fire Protection System Piping Diagram ALL D-02057 Sh. 2B,Plant Fire Protection System Piping Diagram ALL D-02058 Sh. 3B,Plant Fire Protection System Piping Diagram ALL BSEP LAR Rev 2 Page A-46

CP&L Attachment A DBD-62,Water Based Suppression System Sections 3.3.16 & 3.3.31 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.14 [Water Supply Control Valve Supervision] Chapter 3 Requirement: 3.5.14* All fire protection water supply and fire suppression system control valves shall be under a periodic inspection program and shall be supervised by one of the following methods. (a) Electrical supervision with audible and visual signals in the main control room or other suitable constantly attended location. Compliance Statement Compliance Basis (a) Complies - complies by compliance (a) BSEP does not comply with (a) of this with Section (c) NFPA 805 Chapter 3 element, but instead complies with (b) and (c). Reference Document DoDetals OOP-41,Fire Protection and Well Water System ALL OPT-34.2.5.0,Fire Suppression System Control Valve Position ALL Verification OAP-013,Plant Equipment Control ALL Chapter 3 Requirement: (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel. Compliance Statement Compliance Basis (b) Complies - complies by compliance (b) No Additional Clarification. with Section (c) Reference Document Doc Details 0OP-41,Fire Protection and Well Water System ALL OPT-34.2.5.0,Fire Suppression System Control Valve Position ALL Verification OAP-013,Plant Equipment Control ALL Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1 (3) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 Chapter 3 Requirement: (c) Sealing valves in their normal positions. This option shall be utilized only where valves are located within fenced areas or under the direct control of the owner/operator. Compliance Statement Compliance Basis (c) Complies (c) No Additional Clarification. BSEP LAR Rev 2 Page A-47

CP&L Attachment A Reference Document D 0OP-41,Fire Protection and Well Water System ALL 0PT-34.2.5.0,Fire Suppression System Control Valve Position ALL Verification 0AP-013,Plant Equipment Control ALL Fire Protection Safety Evaluation Report, Fire Protection Safety Section 4.3.1 (3) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.15 [Water Supply Hydrant Code Requirements] Chapter 3 Requirement: 3.5.15 Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be provided at intervals of not more than 1000 ft (305 m) along the yard main system. Exception: Mobile means of providing hose and associated equipment, such as hose carts or trucks, shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to the equipment supplied by three hose houses. Compliance Statement Compliance Basis Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC APPROVAL: Per the Fire Protection Program Review (01-01-1977):

                                                "The hydrants on the plant loop are located within approximately 250 feet of the adjacent hydrants. Each hydrant is supplied by a lateral from the yard main, and each lateral is provided with a key-operated (curb) valve."

Per the November 23, 1977 NRC SER, "Yard fire hydrants have been provided at approximately 250 foot intervals around the exterior of the plant.The licensee proposes to extend the fire loop to the service water intake structure to supply sprinklers and manual hose stations in this building. Two new hydrants will be provided on this extension of the fire loop improving coverage in this area.We conclude that, subject to implementation of the above described modifications, the fire water piping system conforms to the BSEP LAR Rev 2 Page A-48

CP&L Attachment A provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable." Per DBD-62:

                                                "In association with upgrades for the Service Water Intake Structure, a nearby yard hydrant will be installed (ref. 6.6.1.2, Section 5.7.6). This was accomplished via PM 77-350 (ref. 6.2.1.4)."

BSEP complies via the exception by providing mobile equipment carts in the fire house and alternate fire equipment building in lieu of hydrant hose house. The mobile equipment cart is described in OFPP-008, and contains fire fighting equipment necessary to support the fire brigade in response to a fire. An inventory to verify that appropriate mobile equipment cart equipment is available for fire brigade use is performed in accordance with 0FPP-008. Reference Document norntails DBD-62,Water Based Suppression System Section 3.3.9 0FP-1017,Code Compliance Evaluation NFPA 24 Code Section 5601 0AP-033,Fire Protection Program Manual Section 5.4.3.2 Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1 (3) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.e.2 Page 9 Plant, January 1, 1977 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.16 [Water Supply Dedicated Limits] Chapter 3 Requirement: 3.5.16* The fire protection water supply system shall be dedicated for fire protection use only. Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis. Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire BSEP LAR Rev 2 Page A-49

CP&L Attachment A protection demand for the specified duration as determined in this section. Compliance Statement Compliance Basis License Amendment Required Per DBD-62, BNP utilizes both exceptions See Attachment L. to NFPA 805 Section 3.5.16:

                                         "1.2.1 SUPPORTING SYSTEMS 1.2.1.2 Demineralized Water System Fire water can be used to fill the SLC

[Standby Liquid Control] Test Tank as a means of alternate coolant injection. This involves using water from the Fire Water Storage Tank, one of the fire pumps, and providing water through one of the fire hose stations on El. 80'-0". The Demineralized Water System provides a secondary water supply for the WFSS [Water-based Fire Suppression System]. Of the 200,000-gallon nominal capacity of the Demineralized Water Tank, 90,000 gallons are reserved specifically for fire suppression (ref. 6.2.8.1, Section 1.3.1.3; ref. 6.1.1.2, Section 9.5.1.4.1.4). 1.2.2 SUPPORTED SYSTEMS 1.2.2.1 Containment Heat Removal In the event that nuclear service water is lost to the RHR [Residual Heat Removal] heat exchangers, the WFSS may be used to provide backup cooling for containment heat removal. This is described in OAOP-18.0 (ref. 6.2.8.4). 1.2.2.2 Coolant Injection In the event of a failure of the normal reactor level control systems to maintain water level, the WFSS may be used as an alternate coolant injection system as follows: Fire water can be used to fill the SLC [Standby Liquid Control] Test Tank as a means of alternate coolant injection. This involves using water from the Fire Water Storage Tank, one of the fire pumps, and providing water through one of the fire hose stations on El. 80'- 0" east of the Reactor Buildings. Fire water can also be used as an alternate injection source for direct injection through a connection between the Well Water flushing line and the Service Water System. This involves using water from the Fire Water Storage Tank, one of the fire pumps, opening 2-FP-PIV-20, and the Well Water flushing BSEP LAR Rev 2 Page A-50

CP&L Attachment A lines to the Service Water System. Fire water can be used for boron injection as follows: (1) filling the RWCU [Reactor Water Cleanup System] Precoat Tank by providing water through one of the fire hose stations on El. 80'-0" east of the Reactor Buildings. This involves using water from the Fire Water Storage Tank, one of the fire pumps, and providing water through one of the fire hose stations on El. 80'-0" east of the Reactor Buildings; or (2) filling the CFD [Condensate Filter Demineralizer] Precoat Tank by providing water through one of the fire hose stations on El. 23'-0" of the Radwaste Building. This involves using water from the Fire Water Storage Tank, one of the fire pumps, and providing water through one of the fire hose stations on El. 23'-0" of the Radwaste Building. 1.2.2.3 Fuel Pool Cooling Fire hoses on the Reactor Building 117' elevation may be used as a makeup water source if the spent fuel pool level cannot by recovered by normal means. This is described in OAOP-38.0 (ref. 6.2.8.2). See Attachment L for further details on the request for NRC approval for non-fire protection uses of the fire protection water supply system. Reference Document Doc Detils DBD-62,Water Based Suppression System Section 1.2.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6 Standpipe and Hose Stations. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. BSEP LAR Rev 2 Page A-51

CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.1 [Standpipe and Hose Station Code Requirements] Chapter 3 Requirement: 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems. Compliance Statement Compliance Basis Complies via previous NRC Approval COMPLIES VIA PREVIOUS NRC APPROVAL: In the original fire protection Complies via EEEE submittal, BSEP stated: CP&L will comply with this requirement [for interior manual hose stations] on the completion of the following items:

a. Water hose stations will be provided in the Control Building, Diesel Generator Building and Service Water Intake Structure as required. Additionally, a water spray system will also be added to the Diesel Generator Building.

b. Additional 2 1/2" hose connections will be provided in the Reactor and AOG Buildings. Additional racks will be provided to cover all areas by a hose stream.

c. No hose stations will be provided inside the reactor containment drywell except that during maintenance and repair periods, temporary fire extinguishers and a fire alarm system will be provided.

d. For certain areas were hose racks are located outside the fire area to avoid possible damage from ruptures or careless operation, hose racks with 100 feet of hose may be utilized to achieve proper coverage.

Per Fire Protection Safety Evaluation Report, Brunswick Nuclear Plant, November 22, 1977:

                                             "(4) Interior Hose Stations Interior hose stations equipped with 1 1/2-inch fire hose have been provided through out the plant; however, some areas are beyond the reach of interior hose streams.

BSEP LAR Rev 2 P age A-52

CP&L Attachment A The licensee proposes to add hose stations in these areas which include the reactor building, the diesel generator building, service water intake structure, control building and augmented offgas building. The hose at some of the interior hose stations is made of unlined linen, which is unsuitable for industrial application. The licensee has proposed to replace this with lined fire hose, and to replace hose racks with equipment suitable for lined fire hose storage. The water supply to hose stations inside the reactor buildings is controlled by a normally closed post-indicator valve in the yard. Before the hose can be used, personnel must be dispatched to open the valve. This could result in a significant delay in use of this equipment for fire fighting. The licensee has proposed to provide a valve which can be operated from the control room or by detectors in the area of the fire to avoid the delay. The nozzles on the interior hose lines are of the adjustable type, approved for use on live electrical equipment. We conclude that, subject to the implementation of these changes, the interior hose installation is acceptable." Per DBD-62, these hose stations and hose connections were installed by PMs 77-347, 77-348, 77-349, 77-353, and 77-354. COMPLIES VIA EEEE: BSEP complies with NFPA 14 as evaluated in OFP-1025, OFP-1026, OFP-1027, 0FP-1028, OFP-1029, and OFP-1030. Reference Document DoDeals APCSB 9.5-1,Fire Protection Program Review, Brunswick Nuclear Section IV.C.3.e.3-5 Plant, January 1, 1977 Fire Protection Safety Evaluation Report,Fire Protection Safety Section 4.3.1(4) Evaluation Report, Brunswick Nuclear Plant, November 22, 1977 OFP-1025,NFPA 14 Code Compliance Engineering Evaluation - ALL Turbine Buildings OFP-1026,NFPA 14 Code Compliance Engineering Evaluation - Units ALL 1 and 2 Reactor Buildings BSEP LAR Rev 2 Page A-53

CP&L Attachment A OFP-1027,Code Compliance Evaluation NFPA 14 - Control Building ALL OFP-1028,NFPA 14 Code Compliance Engineering Evaluation - ALL Radwaste Building OFP-1029,Code Compliance Evaluation NFPA 14- Service Water ALL Building 0FP-1030,Code Compliance Evaluation NFPA 14- Diesel Generator ALL Building DBD-62,Water Based Suppression System Section 3.3.17 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.2 [Standpipe and Hose Station Capability Limitations] Chapter 3 Requirement: 3.6.2 A capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel. Compliance Statement Compliance Basis Complies with Clarification See Implementation Item pertinent to NFPA 805 Chapter 3, Section 3.6.2 compliance in Attachment S of the Transition Report. Reference Document DoDetals 0PT-34.7.2.1,Hose Station Flow ALL 0PLP-01.2,Fire Protection System Operability, Action, and Section 6.4.3 Surveillance Requirements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.3 [Standpipe and Hose Station Nozzle Restrictions] Chapter 3 Requirement: 3.6.3 The proper type of hose nozzle to be supplied to each power block area shall be based on the area fire hazards. The usual combination spray/straight stream nozzle shall not be used in areas where the straight stream can cause unacceptable damage or present an electrical hazard to fire-fighting personnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed. Compliance Statement Compliance Basis BSEP LAR Rev 2 Page A-54

CP&L Attachment A Complies No Additional Clarification. Reference Document Doc hlli DBD-62,Water Based Suppression System Section 3.3.18 UFSAR,Updated Final Safety Analysis Report Section 9.5.1.4.1.4.3.a.1) Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.4 [Standpipe and Hose Station Earthquake Provisions] Chapter 3 Requirement: 3.6.4 Provisions shall be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE). Compliance Statement Compliance Basis Complies with Clarification Seismic standpipes are not an original commitment for BNP. The Federal Register notice that promulgated adoption of NFPA 805 makes the following statement:

                                                  "A commenter noted that Appendix A to BTP APCSB 9.5-1 did not require seismically qualified standpipes and hose stations for operating plants and plants with construction permits issued prior to July 1, 1976. NRC agrees that Appendix A to BTP APCSB 9.5-1 made separate provisions for operating plants and plants with construction permits issued prior to July 1, 1976, and did not require seismically qualified standpipes and hose stations for those plants. Therefore, the requirement in Section 3.6.4 of NFPA 805 is not applicable to licensees with nonseismic standpipes and hose stations previously approved in accordance with Appendix A to BTP APCSB 9.5-1."

Reference Document 66 FR 33356,Final Rule - NFPA 805 Page 33544 Table B-i1 NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A-55

CP&L Attachment A Chapter 3

Reference:

3.6.5 [Standpipe and Hose Station Seismic Connection Limitations] Chapter 3 Requirement: 3.6.5 Where the seismic required hose stations are cross-connected to essential seismic non-fire protection water supply systems, the fire flow shall not degrade the essential water system requirement. Compliance Statement Compliance Basis N/A There are no seismic required hose stations at BNP. Seismic standpipes are not an original commitment for BNP. The Federal Register notice that promulgated adoption of NFPA 805 makes the following statement:

                                            "A commenter noted that Appendix A to BTP APCSB 9.5-1 did not require seismically qualified standpipes and hose stations for operating plants and plants with construction permits issued prior to July 1, 1976. NRC agrees that Appendix A to BTP APCSB 9.5-1 made separate provisions for operating plants and plants with construction permits issued prior to July 1, 1976, and did not require seismically qualified standpipes and hose stations for those plants. Therefore, the requirement in Section 3.6.4 of NFPA 805 is not applicable to licensees with nonseismic standpipes and hose stations previously approved in accordance with Appendix A to BTP APCSB 9.5-1."

Table B-I NFPA 805 Ch.3 Transition Details Chapter 3 References 3.7 Fire Extinguishers. Chapter 3 Requirement: 3.7 Fire Extinguishers. Where provided, fire extinguishers of the appropriate number, size, and type shall be provided in accordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions. Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 10 as BSEP LAR Rev 2 Page A-56

CP&L Attachment A evaluated in OFP-0085. Reference Document Doc Details OFP-0085,Code Compliance Evaluation NFPA 10, Portable ALL Extinguishers Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8 Fire Alarm and Detection Systems. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.1 Fire Alarm. Chapter 3 Requirement 3.8.1 Fire Alarm. Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code. Alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnel assigned to the proprietary alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted: (1) Actuation of any fire detection device Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 72 as evaluated in the applicable portions of (1) Complies 0FP-1031, 0FP-1032, 0FP-1033, 0FP-1035, and 0FP-1036. (1) No Additional Clarification. Reference Document DocDetails Specification 070-011 ,Fire Detection System Sections 2.1.2, 4.1.1.7.1-4.1.1.7.3, 4.2.1.4 0FP-1031,Code Compliance Evaluation NFPA 72E - DG Halon ALL BSEP LAR Rev 2 Page A-57

CP&L Attachment A System and Control Building AO Rooms 0FP-1032,Code Compliance Evaluation NFPA 72E - Turbine ALL Buildings and Transformer Yard 0FP-1033,Code Compliance Evaluation NFPA 72E - Control Building ALL OFP-1036,Code Compliance Evaluation NFPA 72E - Units 1 & 2 ALL Reactor Buildings 0FP-1043,Code Compliance Evaluation NFPA 72D ALL OFP-1035,Code Compliance Evaluation NFPA 72E - Diesel ALL Generator Building and Service Waster Building Chapter 3 Requirement: (2) Actuation of any fixed fire suppression system Compliance Statement Compliance Basis (2) Complies (2) No Additional Clarification. Reference Document DoDetals Specification 070-011, Fire Detection System Section 4.1.1.7.1 Chapter 3 Requirement: (3) Actuation of any manual fire alarm station Compliance Statement Compliance Basis (3) Complies (3) No Additional Clarification. Reference Document Doc Details Specification 070-011 ,Fire Detection System Section 4.1.1.7.1 Chapter 3 Requirement: (4) Starting of any fire pump Compliance Statement Compliance Basis (4) Complies (4) No Additional Clarification. Reference Document Doc Details Specification 070-011,Fire Detection System Section 4.2.12.1 Chapter 3 Requirement: (5) Actuation of any fire protection supervisory device Compliance Statement Compliance Basis (5) Complies (5) No Additional Clarification. Reference Document Doc 4Dtals1 Specification 070-011 ,Fire Detection System Section 4.3.1.2 BSEP LAR Rev 2 Page A-58

CP&L Attachment A Chapter 3 Requirement: (6) Indication of alarm system trouble condition Compliance Statement Compliance Bases (6) Complies (6) No Additional Clarification. Reference Document DoeDetails Specification 070-011 ,Fire Detection System Section 4.1.1.7.3 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.1.1 [Fire Alarm Communication Requirements] Chapter 3 Requirement: 3.8.1.1 Means shall be provided to allow a person observing a fire at any location in the plant to quickly and reliably communicate to the control room or other suitable constantly attended location. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DocDDtails 0PFP-013,General Fire Plan Section 3.1.1 OPLP-01.2,Fire Protection System Operability, Action, and Section 6.9 Surveillance Requirements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.1.2 [Fire Alarm Prompt Notification Limits] Chapter 3 Requirement: 3.8.1.2 Means shall be provided to promptly notify the following of any fire emergency in such a way as to allow them to determine an appropriate course of action: (1) General site population in all occupied areas. Compliance Statement Compliance Basis (1) Complies (1) No Additional Clarification Reference Document DocoDtails 0AP-033,Fire Protection Program Manual Section 5.6.1 BSEP LAR Rev 2 Page A-59

CP&L Attachment A OPLP-01.2,Fire Protection System Operability, Action, and Section 6.9 Surveillance Requirements Chapter 3 Requirement: (2) Members of the industrial fire brigade and other groups supporting fire emergency response Compliance Statement Compliance Basis (2) Complies (2) No Additional Clarification Reference Document Doc Details OPFP-013,General Fire Plan Section 2.1 & 3.2 OPLP-01.2,Fire Protection System Operability, Action, and Section 6.9 Surveillance Requirements Chapter 3 Requirement: (3) Off-site fire emergency response agencies. Two independent means shall be available (e.g., telephone and radio) for notification of off-site emergency services Compliance Statement Compliance Basis (3) Complies (3) No Additional Clarification Reference Document Doc Details UFSAR,Updated Final Safety Analysis Report Section 9.5.2.2 OERP,Radiological Emergency Response Plan Section A.1.4 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.2 Detection. Chapter 3 Requirement: 3.8.2 Detection. If automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes. Compliance Statement Compliance Basis Complies via EEEE BSEP complies with NFPA 72 as evaluated in the applicable portions of OFP-1031, OFP-1032, OFP-1033, OFP-1035, and OFP-1036. Reference Document Dc Detals~ 0FP-1031,Code Compliance Evaluation NFPA 72E - DG Halon ALL System and Control Building AO Rooms BSEP LAR Rev 2 Page A-60

CP&L Attachment A OFP-1032,Code Compliance Evaluation NFPA 72E - Turbine ALL Buildings and Transformer Yard 0FP-1033,Code Compliance Evaluation NFPA 72E - Control Building ALL OFP-1036,Code Compliance Evaluation NFPA 72E - Units 1 & 2 ALL Reactor Buildings OFP-1043,Code Compliance Evaluation NFPA 72D ALL 0FP-1035,Code Compliance Evaluation NFPA 72E - Diesel ALL Generator Building and Service Waster Building BNP-0160,Table B Fire Area Transition ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9 Automatic and Manual Water-Based Fire Suppression Systems. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.1 [Fire Suppression System Code Requirements] Chapter 3 Requirement: 3.9.1

• If an automatic or manual water-based fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including the following:

(1) NFPA 13, Standard for the Installation of Sprinkler Systems Compliance Statement Compliance Basis (1) Complies via EEEE (1) COMPLIES VIA EEEE: BSEP complies with NFPA 13 as evaluated in the Complies with Clarification applicable portions of OFP-1038, 0FP-1039, 0FP-1041, and OFP-1042. COMPLIES WITH CLARIFICATION: The water-based fire suppression systems required to meet the performance or deterministic requirements of Chapter 4 meet a standard of design, construction, BSEP LAR Rev 2 Page A-61

CP&L Attachment A maintenance, inspection and testing that is consistent with the applicable NFPA code(s). Since the majority of the required Systems at BSEP were designed and installed per applicable NFPA codes, a specific NFPA Code Compliance Review is not the only means of showing compliance. Only minor deviations from code requirements were found with the majority accepted "as-is" via engineering evaluation. There is reasonable assurance of this conformance by virtue of the methodology normally used in the industry to control the design conformance, quality and ongoing performance of all NFPA code systems. It does not appear that expansion of these calculations for additional systems, many of which have been installed since original construction, would provide significant added benefit or safety beyond that currently in place. Original system design and installation was monitored by detailed specification development and adherence and internal quality assurance/control programs, along with review and approval by outside insurance underwriters. The plant modification process controls what changes can be made to insure that the code requirements are maintained. Internal BSEP programs such as the Engineering Program self-assessments and System Engineering monitoring and trending efforts provide continuous oversight of the systems to ensure their design and performance are maintained. These aspects, in combination with original plant construction and on-going system maintenance, provide assurance that the systems continue to meet the original NFPA code requirements and provide a suitable approach for demonstrating compliance with the NFPA 805 Chapter 3 requirements for systems determined to be required. Reference Document Doc Details OFP-1038,Code Compliance Evaluation NFPA 13- Reactor Buildings ALL OFP-1039,Code Compliance Evaluation NFPA 13- Control Building ALL BSEP LAR Rev 2 Page A-62

CP&L Attachment A OFP-1041,Code Compliance Evaluation NFPA 13- Service Water ALL Building OFP-1042,Code Compliance Evaluation NFPA 13- Diesel Generator ALL Building BNP-0160,Table B Fire Area Transition ALL Chapter 3 Requirement: (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection Compliance Statement Compliance Basis (2) Complies via EEEE (2) BSEP complies with NFPA 15 as evaluated in the applicable portions of OFP- 1024. Reference Document neta"Is OFP-1024,Unit 1 & 2 Reactor Buildings NFPA 15 Code Compliance ALL Engineering Evaluation Chapter 3 Requirement: (3) NFPA 750, Standard on Water Mist Fire Protection Systems Compliance Statement Compliance Basis (3) N/A (3) Water mist fire protection systems are not used at BSEP. Reference Document DoDetals DBD-62,Water Based Suppression System Section 0.1.4 Chapter 3 Requirement: (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems Compliance Statement Compliance Ragsi (4) N/A (4) No Foam-Water Sprinkler or Foam-Water Spray systems are required to meet the performance or deterministic requirements of Chapter 4. Reference Document DoDetals BNP-0160,Table B Fire Area Transition ALL UFSAR,Updated Final Safety Analysis Report Section 9.5.1.5 Table B-1 NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A--63

CP&L Attachment A Chapter 3 Referencee 3.9.2 [Fire Suppression System Flow Alarm] Chapter 3 Requirement: 3.9.2 Each system shall be equipped with a water flow alarm. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DBD-62,Water Based Suppression System Sections 0.1.4.1, 0.1.4.2, & 3.1.4 Specification 070-011 ,Fire Detection System Attachments A & Attachment 4 D-02303-SH0001,Diesel Generator Building Fire Protection Sprinkler ALL System Piping Diagram D-29099-SH0001,Reactor Building Fire Protection Piping Sprinkler ALL System Piping Diagram D-29099-SH0002,Reactor Building Piping Diagram Fire Protection ALL Piping Diagram Fire Protection Piping Sprinkler D-02299,Reactor Building Fire Protection Piping Sprinkler System ALL Piping Diagram D-02058-SH003B,Plant Fire Protection System Piping Diagram ALL D-02304,Piping Diagram Service Water Radwaste & Treatment ALL Buildings Fire Protection Sprinkler System Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.3 [Fire Suppression System Alarm Locations] Chapter 3 Requirement: 3.9.3 All alarms from fire suppression systems shall annunciate in the control room or other suitable constantly attended location. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DoDetails Specification 070-011 ,Fire Detection System Sections 4.1.1.7.1 & 4.1.1.7.3 DBD-62,Water Based Suppression System Section 3.1.1 Table B-1 NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A-64

CP&L Attachment A Chapter 3

Reference:

3.9.4 [Fire Suppression System Diesel Pump Sprinkler Protection] Chapter 3 Requirement: 3.9.4 Diesel-driven fire pumps shall be protected by automatic sprinklers. Compliance Statement Compliance Bases Complies No Additional Clarification. Reference Document no.Dntails DBD-62,Water Based Suppression System Section 3.3.47 F-03568,Water Treatment Building General Arrangement and ALL Grounding Plan 0-FP-20563,Fire Protection Piping Isometric, Water Treatment ALL Building Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.5 [Fire Suppression System Shutoff Controls] Chapter 3 Requirement: 3.9.5 Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Documentc Detail DBD-62,Water Based Suppression System Section 3.3.31 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.6 [Fire Suppression System Valve Supervision] Chapter 3 Requirement: 3.9.6 All valves controlling water-based fire suppression systems required to meet the performance or deterministic requirements of Chapter 4 shall be supervised as described in 3.5.14. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details BSEP LAR Rev 2 Page A-65

CP&L Attachment A 00P-41 ,Fire Protection and Well Water System ALL 0PT-34.2.5.0, Fire Suppression System Control Valve Position ALL Verification 0AP-013,Plant Equipment Control ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10 Gaseous Fire Suppression Systems. Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.1 [Gaseous Suppression System Code Requirements] Chapter 3 Requirement: 3.10.1 If an automatic total flooding and local application gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes: (1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems Compliance Statement Compliance Basis (1) Complies via EEEE (1) BSEP complies with NFPA 12 as evaluated in the applicable portions of OFP-1019. Reference Document DoDetals 0FP-1019,Code Compliance Evaluation NFPA 12 ALL Chapter 3 Requirement: (2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems Compliance Statement Compliance Basis (2) Complies via EEEE (2) BSEP complies with NFPA 12A as evaluated in the applicable portions of 2FP-1015. BSEP LAR Rev 2 Page A-66

CP&L Attachment A Reference Document 2FP-1015,NFPA 12A Code Compliance Engineering Evaluation ALL Chapter 3 Requirement: (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems Compliance Statement Compliance Basis (3) N/A (3) BSEP does not utilize any Clean Agent Fire Extinguishing Systems Reference Document DDetals DBD-61,Gaseous Suppression System Section 0.2.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.2 [Gaseous Suppression System Alarm Location] Chapter 3 Requirement: 3.10.2 Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other constantly attended location identified. Compliance Statement Compliance Bases Complies No Additional Clarification. Reference Document DocDetals DBD-61,Gaseous Suppression System Sections 3.1.1 & 3.1.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3 References 3.10.3 [Gaseous Suppression System Ventilation Limitations] Chapter 3 Requirement: 3.10.3 Ventilation system design shall take into account prevention from over-pressurization during agent injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants. Compliance Statement Compliance Basis Complies via EEEE BSEP complies with the venting requirements as evaluated in the applicable portions of 2FP-1015 and OFP-1019. BSEP LAR Rev 2 Page A-67

CP&L Attachment A Reference Document Doc Details 0FP-1019,Code Compliance Evaluation NFPA 12 Attachment 4 (Code Section 26), Summary Section

1) & 2)

SD-37,Reactor Building HVAC Section 1.4.1 DBD-61 ,Gaseous Suppression System Section 3.3.1 2FP-1015,NFPA 12A Code Compliance Engineering Evaluation Code Section 2-2.2.3 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.4 [Gaseous Suppression System Single Failure Limits] Chapter 3 Requirement: 3.10.4* In any area required to be protected by both primary and backup gaseous fire suppression systems, a single active failure or a crack in any pipe in the fire suppression system shall not impair both the primary and backup fire suppression capability. Compliance Statement Compliance Basis N/A No areas at BSEP are protected by both a primary and backup gaseous fire suppression system. Reference Document Doc Details UFSAR,Updated Final Safety Analysis Report Section 9.5.1.5 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.5 [Gaseous Suppression System Disarming Controls] Chapter 3 Requirement: 3.10.5 Provisions for locally disarming automatic gaseous suppression systems shall be secured and under strict administrative control. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DocDetails OAP-033,Fire Protection Program Manual Section 5.3.4.7 OPS-NGGC-1301 ,Equipment Clearance ALL 001-01.08,Control of Equipment and System Status ALL BSEP LAR Rev 2 Page A-68

CP&L Attachment A CP&L Attachment A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.6 [Gaseous Suppression System C02 Limitations] Chapter 3 Requirement: 3.10.6* Total flooding carbon dioxide systems shall not be used in normally occupied areas. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document DoDetaeIs BNP-0124,BNP Fire Compartment Transient Fire Influencing Factors Attachment A OOP-41,Fire Protection and Well Water System Section 3.7 OPS-NGGC-1301,Equipment Clearance ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.7 [Gaseous Suppression System C02 Warnings] Chapter 3 Requirement: 3.10.7 Automatic total flooding carbon dioxide systems shall be equipped with an audible pre-discharge alarm and discharge delay sufficient to permit egress of personnel. The carbon dioxide system shall be provided with an odorizer. Compliance Statement Compliance Basis Complies with Clarification See Implementation Item pertinent to NFPA 805 Chapter 3, Section 3.10.7 compliance in Attachment S of the Transition Report. Reference Document DocDetails DBD-61 ,Gaseous Suppression System Section 3.1.1 OFP-1019,Code Compliance Evaluation NFPA 12 Summary Section Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.8 [Gaseous Suppression System C02 Required Disarming] BSEP LAR Rev 2 Page A-69

CP&L Attachment A Chapter 3 Requirement: 3.10.8 Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during work in the protected space. Compliance Statement Complioance Basis Complies No Additional Clarification. Reference Document OOP-41,Fire Protection and Well Water System Section 3.7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.9 [Gaseous Suppression System Cooling Considerations] Chapter 3 Requirement: 3.10.9 The possibility of secondary thermal shock (cooling) damage shall be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide. Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details DBD-61,Gaseous Suppression System Section 3.3.1 SFPE Handbook of Fire Protection Engineering,SFPE Handbook of Page 4-125 Fire Protection Engineering, Fourth Edition BNP-0230,Change No. BNP-0230, Rev. 0 ALL OFP-1019,Code Compliance Evaluation NFPA 12 ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.10 [Gaseous Suppression System Decomposition Issues] Chapter 3 Requirement: 3.10.10 Particular attention shall be given to corrosive characteristics of agent decomposition products on safety systems. Compliance Statement Compliance Basis Complies No Additional Clarification. Reference Document Doc Details BSEP LAR Rev 2 Page A-70

CP&L Attachment A DBD-61,Gaseous Suppression System ALL NFPA Fire Protection Handbook,NFPA Fire Protection Handbook, Sections 17.1 & 17.6 20th Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11 Passive Fire Protection Features Chapter 3 Requirement: 3.11 Passive Fire Protection Features. This section shall be used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire. Compliance Statement Compliance Basis N/A N/A - Section Heading, no technical requirements. See sub-sections for specific compliance statements and references. Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.1 Building Separation. Chapter 3 Requirement: 3.11.1 Building Separation. Each major building within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 hours or by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, Recommended Practice for Protection of Buildings from Exterior Fire Exposures. Exception: Where a performance-based analysis determines the adequacy of building separation, the requirements of 3.11.1 shall not apply. Compliance Statement Compliance Aasl i Complies No Additional Clarification. Reference Document DocDetals~ UFSAR,Updated Final Safety Analysis Report Section 9.5.1.5 F-04001,Sanitary & Storm Sewers, Well Water, Potable Water, and ALL Fire Protection Piping Plot Plan EC 68540,lnstall VFD's and Associated Heat Exchangers Section B.4.24 BSEP LAR Rev 2 Page A-71

CP&L Attachment A OFP-1206,Evaluation of the Spatial Separation between the Turbine Building and Power Distribution Centers Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.2 Fire Barriers. Chapter 3 Requirement: 3.11.2 Fire Barriers. Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be designed and installed to meet the specific fire resistance rating using assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Fire Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials. Compliance Statement Comp*iance Basis Complies COMPLIES: No Additional Clarification Complies via EEEE Reference Document Doc Details UFSAR,Updated Final Safety Analysis Report Section 9.5.1.4.3.2 0FP-0031,Evaluation of Joint Seals in the DGB Stairwells and in the ALL Control Room 0FP-1007,Evaluation of Battery Room 1A (CB- 07) and Battery Room ALL 1B (CB-08) Joint Seal (Penetration CB-1-485) 2FP-1013,Evaluation of Switchgear Room E3 and Switchgear Room ALL E4 Joint Seal Penetration (DG-3-443) 2FP-1014,Evaluation of Penetration Seal CB-1-488 ALL OFP-0023,Evaluation of Fire Barrier Penetrations T1 016 & T2 ALL 016, rev. 0 OFP-0028,Service Water Bldg. Block Wall Gap Seal, Rev. 0 ALL OFP-0032,Embedded Combustible Material - Turbine Building ALL OFP-0033,Structural Steel Fireproofing ALL OFP-0035,Embedded Combustible Material Resin Storage Room, ALL Rev. 0 2FP-1009,Evaluation of Penetration Seal DG-3-448 ALL 704U-M-17,Downgrade of East Wall of SWIS Between El. (-) 8.63 & ALL 20.0 and Partial Floor at 20.0", Rev. 2 85-125-0-47-F,D.G. Building Spare Cable - North Wall ALL 87-0301 ,Steel Plate Wall Evaluation ALL 89-0094,Revision to Rodofoam Seal to Provide Fire Rating ALL 92-0169,Evaluation for Acceptability of Gap Seal ALL BSEP LAR Rev 2 Page A-72

CP&L Attachment A 99-00428,Evaluate 2-FP-DG-2-340 for New Seal Design ALL OFP-0037,Exposed Rebar in Fire Barriers ALL 704U-M-24,AOG Building Concrete Structures Evaluation ALL 85-0186,Diesel Generator Pedestal Seal; Rodofoam Evaluation ALL 85-125-0-33-F, Inaccessible Fire Barriers ALL 90-0286,Downgrades Control Room Floor ALL OFP-0006,Acceptance Criteria for Block Wall Expansion ALL 89-0010,Evaluate Fire Hazard of Existing Rodofoam 300 Used as ALL Seismic Gap Filler Table B-i NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.3 Fire Barrier Penetrations. Chapter 3 Requirement: 3.11.3* Fire Barrier Penetrations. Penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by Chapter 4. (See 3.11.3.4 for penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable: (1) NFPA 80, Standard for Fire Doors and Fire Windows. Compliance Statement Compliance Basis (1) Complies via EEEE (1) BSEP complies with NFPA 80 as evaluated in ESR 97-00571. Reference Document DLDetals ESR 97-00571,Fire Door Problem Resolution ALL FAQ 06-0020,Identification of "applicable NFPA standards" ALL NFPA 101,Life Safety Code, 2003 Edition Sections 8.3.3.2.1(a) & 9.2.1 OFP-0091,Code Compliance Evaluation NFPA 90A, Code ALL Compliance Evaluation for NFPA 90A, Installation of Air Conditioning and Ventilating Systems - 1985 Edition OFP-1058,Evaluation of the Equipment Hatch Located on El. 23'-0" of ALL the Diesel Generator Building used to Cover Penetration Seal 2-FP-DG-2-025 90-0139,Evaluation of a Steel Plate Covering Penetration CB-1-066 ALL in Battery Room 1A 94-00793,Evaluation of the Unit 1 and 2 Diesel Generator Building ALL Equipment Hatches Located on El. 50'-0" of the Diesel Generator

Building, BSEP LAR Rev 2 Page A-73

CP&L Attachment A 704U-M-33,AOG Building Penetration No. AO-2-032 Evaluation ALL 84-0615,Penetration Evaluation - SWIS West Wall ALL 88-0449,Evaluation of Missing Thru-Bolt in Diesel Cell #4 Fire Barrier ALL 85-125-0-54-F,Abandoned Hilti Bolt Shells in Concrete or Masonry ALL Fire Barriers or Uncontained Holes in Concrete Fire Barriers. Chapter 3 Requirement: (2) NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating Systems. Compliance Statement Compliance Basis (2) Complies via EEEE (2) BSEP complies with NFPA 90A as evaluated in BSEP Calculation 0FP-0091. Reference Document DocDetals ESR 97-00571 ,Fire Door Problem Resolution ALL FAQ 06-0020,Identification of "applicable NFPA standards" ALL NFPA 101,Life Safety Code, 2003 Edition Sections 8.3.3.2.1(a) & 9.2.1 OFP-0091,Code Compliance Evaluation NFPA 90A, Code ALL Compliance Evaluation for NFPA 90A, Installation of Air Conditioning and Ventilating Systems - 1985 Edition OFP-1058,Evaluation of the Equipment Hatch Located on El. 23'-0" of ALL the Diesel Generator Building used to Cover Penetration Seal 2-FP-DG-2-025 90-0139,Evaluation of a Steel Plate Covering Penetration CB-1-066 ALL in Battery Room 1A 94-00793,Evaluation of the Unit 1 and 2 Diesel Generator Building ALL Equipment Hatches Located on El. 50'-0" of the Diesel Generator

Building, 704U-M-33,AOG Building Penetration No. AO-2-032 Evaluation ALL 84-0615,Penetration Evaluation - SWIS West Wall ALL 88-0449,Evaluation of Missing Thru-Bolt in Diesel Cell #4 Fire Barrier ALL 85-125-0-54-F,Abandoned Hilti Bolt Shells in Concrete or Masonry ALL Fire Barriers or Uncontained Holes in Concrete Fire Barriers.

Chapter 3 Requirement: (3) NFPA 101, Life Safety Code Exception: Where fire area boundaries are not wall-to-wall, floor-to-ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall be required to assess the adequacy of fire barrier forming the fire boundary to determine if the barrier will withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to be protected by other means as acceptable to the AHJ. Compliance Statement Compliance Basis (3) Complies with Clarification (3) NFPA 101 is not a committed code for BSEP and as such is not part of the BSEP LAR Rev 2 Page A-74

CP&L Attachment A current licensing basis. Per FAQ 06-0020 the following guidance applies as to which NFPA standards referenced in Chapter 3 are applicable:

                                              "Where used in NFPA 805, Chapter 3, the term, "applicable NFPA Standards" is considered to be equivalent to those NFPA standards identified in the current license basis (CLB) for procedures and systems in the Fire Protection Program that are transitioning to NFPA 805. New Fire Protection Systems would be subject to the most current code or standard."

NFPA 101, Section 8.2.3.2.1(a) with regards to rated fire door assemblies refers to NFPA 80. NFPA 101 Section 9.2.1 with regards to rate fire dampers refers to NFPA 90A. NFPA 101 compliance is achieved through compliance with NFPA 80 and NFPA 90A as described in sections (1) and (2) of this element. Reference Document DLDetails ESR 97-00571,Fire Door Problem Resolution ALL FAQ 06-0020,Identification of "applicable NFPA standards" ALL NFPA 101,Life Safety Code, 2003 Edition Sections 8.3.3.2.1(a) & 9.2.1 OFP-0091,Code Compliance Evaluation NFPA 90A, Code ALL Compliance Evaluation for NFPA 90A, Installation of Air Conditioning and Ventilating Systems - 1985 Edition OFP-1058,Evaluation of the Equipment Hatch Located on El. 23'-0" of ALL the Diesel Generator Building used to Cover Penetration Seal 2-FP-DG-2-025 90-0139,Evaluation of a Steel Plate Covering Penetration CB-1-066 ALL in Battery Room 1A 94-00793,Evaluation of the Unit 1 and 2 Diesel Generator Building ALL Equipment Hatches Located on El. 50'-0" of the Diesel Generator

Building, 704U-M-33,AOG Building Penetration No. AO-2-032 Evaluation ALL 84-0615,Penetration Evaluation - SWIS West Wall ALL 88-0449,Evaluation of Missing Thru-Bolt in Diesel Cell #4 Fire Barrier ALL 85-125-0-54-FAbandoned Hilti Bolt Shells in Concrete or Masonry ALL Fire Barriers or Uncontained Holes in Concrete Fire Barriers.

Table B-i NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A-75

CP&L Attachment A Chapter 3

Reference:

3.11.4 Through Penetration Fire Stops. Chapter 3 Requirement: 3.11.4 Through Penetration Fire Stops. Through penetration fire stops for penetrations such as pipes, conduits, bus ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall be protected as follows. (a) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period. Compliance Statement Compliance Basis (a) Complies (a) No Additional Clarification Complies via Previous NRC Approval The Complies via Previous NRC Approval Complies via EEEE compliance strategy refers to the referenced Safety Evaluation, dated May 29, 1987 which stated the following:

                                                "The BSEP acceptance criteria developed for the penetration seals contains the following three major elements:

1. The test conditions will use the standard fire exposure curve as defined in ASTM E-119. This is the same requirement for all three referenced criteria (NRC, ANI, and IEEE);

2. The standard hose-stream test will be conducted as specified in ASTM E-1 19.

This is a stricter test than any of the three referenced criteria in that only a solid stream is allowed by ASTM; and

3. The temperature rise criteria as defined by ANI was selected with one additional consideration. When a recorded temperature exceeds the temperature rise limit of 325F, the situation will be analyzed and can be dispositioned if justified. This criterion is less strict than the NRC limits, but more strict than IEEE.

3.0 CONCLUSION

S After having reviewed the penetration seal program, the staff concludes that the acceptance criteria established by the licensee as well as the various seal installation configurations, are acceptable, and are suitable deviations from BTP ASB 9.5-1. We concur with the licensee that the BSEP LAR Rev 2 Page A-76

CP&L Attachment A additional 75°F temperature rise allowed by the ANI criteria is not considered likely to significantly add to the risk of igniting material on the unexposed side of the barrier. Therefore, there is no need to add to the administrative controls already in place with respect to the control of combustibles inside the plants." The Complies via EEEE compliance strategy refers to the referenced list of evaluations where the various seal configurations were evaluated to be acceptable for the hazards in which they are installed. These various identified penetration seal locations are evaluated in each of the referenced EEEE's. The attached Table identifies the applicable Fire Zones/Fire Areas for each of the EEEE's. Reference Document DocDetails Specification 118-003,Selection and Installation of Fire Barrier and ALL Pressure Boundary Penetration Seals Safety Evaluation,Safety Evaluation, Fire Barrier Penetration Seals, ALL May 29, 1987 NLS-86-448,Silicon Based Sealants Fire Rated Link Seals Fire ALL Rated Dampers & Grout Attachments, 12/31/1986 2FP-0036,Evaluation of the Penetration Seal in the Unit 2 Reactor ALL Building ECCS Room 2FP-0050,Evaluation of Control Building Selected Penetrations ALL 704U-M-28,Evaluation of Seal Design D1 (From Spec Waiver SWB- ALL 118-003-H) When Installed From One Side of a Barrier in the AOG Building OFP-1060,Fire Resistance Rating for Penetration Seals R2-2-021 ALL through R2-2-027 85-125-0-12-F,Evaluation of Control Building Flush Mounted ALL Junction Boxes 85-125-0-16,Evaluation of Diesel Generator Building CGB ALL Connectors 85-125-0-17,Evaluation of Reactor Building PAM Tubing ALL Penetrations, Unit 1 & 2 85-125-0-18,Evaluation of Diesel Generator Building Penetration ALL Seals 85-125-0-27,Evaluation of Bus Duct Seals in the Diesel Generator ALL BSEP LAR Rev 2 Page A-77

CP&L Attachment A Building 85-125-0-32,Evaluation of Control Building Penetrations behind Pull ALL Boxes 85-125-0-34, Evaluation of Penetration Seal 2-FP-R2-3-008 ALL 85-125-0-38,Evaluation of Penetrations 1-FP-R1-4-001 and 2-FP- ALL R2-4-001 85-125-0-42,Evaluation of Control Building Penetration 0-FP-CB ALL 257 85-125-0-45,Sealing Requirements for Penetration 0-FP-CB-2-277 ALL 85-125-0-48,Evaluation of Penetration Seal R2-1-012 ALL 85-125-0-53,Evaluation of Two Conduits in the Unit 2 Reactor ALL Building ECCS Room OFP-0021,Downgrade of Rattle Space Wall-to-Sleeve Link-Seals to ALL Non-Fire Rated Status 0FP-0026,Battery Room Penetration Seals ALL 704U-M-26,Evaluation of the Use of Nelson CLK and RSW for 3 ALL Hour Fire Barrier Penetration Seals for Spare Open Conduits 84-0622,Penetration Evaluation - SWIS East Wall ALL 85-125-0-08-F,Diesel Generator Bldg. Pyrocreted Pull Box ALL Enclosures 85-125-0-14-F,PVC Pipe Penetration SW-3-031 in Service Water ALL Intake Structure 85-125-0-21-F,Evaluation of Seals CB-1-262, 263, 264, 265, 270, ALL 271, 272, & 273 85-125-0-23-F,Turbine Bldg-Combination Link Seal and Additional ALL Moisture Seals 85-125-0-31-F,Unit 1 & 2 Turbine Bldg/Reactor Wall Thru Pipe Link ALL Seals 85-125-0-49-F,Reactor Bldg.- Eccentric Link-Seal Design ALL 85-125-0-51-F, Existing Link Seal Evaluations ALL 89-0149,Elevation of Service Water Building Penetration Seal SW ALL 031 95-00642,Alternative Repair to Fire Barrier Penetration Seals ALL 95-01461,Evaluate Fire Rating of Penetration R2-1-009 ALL 98-00054,Eval. of Silicone Foam Fire Seals Containing Copper Pipe ALL 704U-M-22/S1,Evaluation of Conduit Penetrations in Reactor ALL Buildings and Control Building 85-125-0-41-F,Reactor Bldg.-RHR Rooms Penetration Seals ALL 704U-M-31 ,Sealing Requirements for Penetration AO-2-057 in AOG ALL Bldg 85-125-0-02-F,U/1 & U/2 Reactor Building ECCS Room Minimum ALL Embedment of Hilti Kwik Bolts for Boot Seal Penetration Fire Seals 85-125-0-05-F,Control Buiding Pull/Junction Box Fire Stop ALL BSEP LAR Rev 2 Page A-78

CP&L Attachment A Application 85-125-0-10-F,Diesel Gen. Bldg. Pyrocrete Enclosure Barriers of ALL Pipe & Conduit 85-125-0-11-F,Diesel Generator Building Evaluation of Thermo-Lag ALL Installation 85-125-0-22-F,Deviation to Design "C" of Specification # 118-003 ALL 704U-S-03,12" Diameter Grouted Sleeved Opening, Fire Seal ALL Evaluation 85-125-0-04-F,Cellular Concrete Floor & Wall Blockout Electrical ALL Penetrations Chapter 3 Requirement: (b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and shall be permitted to be installed on either side of the barrier in a location that is as close to the barrier as possible. Exception: Openings inside conduit 4 in. (10.2 cm) or less in diameter shall be sealed at the fire barrier with a fire-rated internal seal unless the conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot gas seal in this application. Compliance Statement Compliance Basis (b) Complies (b) COMPLIES: No Additional Clarification Complies via EEEE Reference Document DoDetals Specification 118-003,Selection and Installation of Fire Barrier and Section 5.3 Pressure Boundary Penetration Seals NED-B/MECH-1001 ,Evaluates Steel Conduit Caps as Rated Seal; ALL Generic to All Plants 85-125-0-06-F,Diesel Generator Bldg. Evaluation of 3M-Interam ALL 85-125-0-07-F,Qualification of Kaowool Conduit Wrap as Equivalent ALL Non-Combustible Seals (as required in Spec 118-003)-DG Bldg. 704U-M-31 ,Sealing Requirements for Penetration AO-2-057 in AOG ALL Bldg 85-125-0-11-F,Diesel Generator Building Evaluation of Thermo-Lag ALL Installation 704U-M-27,Evaluation of Chico A3 and Chico X Noncombustible ALL Seals in Augmented Off-Gas Building 85-125-0-25-F,Evaluation of Penetrations in Control Building, Fire ALL Zone CB-5 85-125-0-28-F,Evaluation of Diesel Generator Building 4160V ALL Switchgear Conduit Internal Seals 85-125-0-46-F,Evaluation of Control Building Cable-To-Conduit Seal ALL BSEP LAR Rev 2 Page A-79

CP&L Attachment A 85-125-0-01-F,Diesel Generator Building - El 50'-0" Penetration ALL Seals Inside Conduit to 4160 V Switchgear 85-125-0-26-F,Control Bldg. Cable Accessway Penetration ALL Evaluation 85-125-0-35-F,Evaluation of Diesel Generator Building 50'-0" Elev. ALL Floor Penetrations Inside 4KV Switchgear 85-125-0-39-F,Evaluation of Diesel Generator Bldg. Penetration No. ALL 5 of DG-2-135 for Conduit 31V1-CB ESS2 85-125-0-40-F,Control Bldg.-Termination Box Conduit Internal Seal ALL 704U-M-22,Evaluation of Conduit Penetrations in Reactor Buildings ALL and Control Building 85-125-0-20-F,Evaluation of Diesel Generator Building CGB ALL Connectors 85-125-0-36-F,Steel Plate/Cap on Pipe Thru Fire Barrier ALL 99-00043,Evaluation of Fire Seal 2-FP-DG-2-026-7 ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.5 Electrical Raceway Fire Barrier Systems (ERFBS). Chapter 3 Requirement: 3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS). ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area. ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area."; The ERFBS needs to adequately address the design requirements and limitations of supports and intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill, and type shall be demonstrated. Exception No. 1: When the temperatures inside the fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Training Within the Same Fire Area," Supplement 1, functionality of the cable at these elevated temperatures shall be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure." Exception No. 2: ERFBS systems employed prior to the issuance of Generic Letter 86-10, Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance. Compliance Statement Compliance Bassi Complies COMPLIES: No Additional Clarification BSEP LAR Rev 2 Page A-80

Attachment A Note - Modifications identified in Table S-1 items 5 and 7 will comply with these TableB1NP80Ch3TastoDeal specified requirements. Table B-i NFPA 805 Ch.3 Transition Details BSEP LAR Rev 2 Page A-81

CP&L Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Table from Section 3.11.4 Evaluation Title Fire Zone Fire Area 2FP-0036 Evaluation of the Penetration Seal in the Unit 2 Reactor RB2-6 RB2-6 Building ECCS Room RB2-01 n RB2-1 2FP-0050 Evaluation of Control Building Selected Penetrations CB-01 b CB-1 CB-5, CB-6, CB-21 CB-23E TB2-01 a TB 1 RW-01 b RW-1 TB1-01b TB1 U2 TB BLDG EAST HALLWAY, Fire Zone not TB1 assigned part of Fire Area TB1 704U-M-28 Evaluation of Seal Design D1 (From Spec Waiver AOG-1 AOG-1 SWB-1 18-003-H) When Installed From One Side of a Barrier in the AOG Building OFP-1 060 Fire Resistance Rating for Penetration Seals R2-2-021 RB2-04 RB2-1 through R2-2-027 TB2-01 b TB1 85-125-0-12-F Evaluation of Control Building Flush Mounted Junction CB-20 CB-23E Boxes TB2-01 a TB1 85-125-0-16-F Evaluation of Junction Box Penetration Seals within a DG-02 DG-2 Floor Slab in the Diesel Generator Bldg DG-03 DG-3 DG-13 DG-13 DG-14 DG-14 85-125-0-17-F Evaluation of Reactor Building PAM Tubing RB1-01g(S/W) RB1-1 Penetrations, Unit 1 & 2 TB1-01a TB1 RB2-01 g(N/W) RB2-1 TB2-01 b TB1 85-125-0-18-F Evaluation of Diesel Generator Building Penetration DG-20 DG-20 Seals DG-21 DG-21 DG-22 DG-22 85-125-0-27-F Evaluation of Bus Duct Seals in the Diesel Generator DG-1 1 DG-11 Building DG-12 DG-12 DG-13 DG-13 DG-14 DG-14 85-125-0-32-F Evaluation of Control Building Penetrations behind Pull CB-20 CB-23E Boxes Ul TB BLDG EAST HALLWAY, Fire Zone not TB1 assigned part of Fire Area TB1 BSEP LAR Rev 2 Page A-82

CP&L Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Table from Section 3.11.4 Evaluation Title Fire Zone Fire Area 85-125-0-34-F Evaluation of Penetration Seal 2-FP-R2-3-008 RB2-01h (W/C) RB2-1 U2 TB BLDG EAST HALLWAY, Fire Zone not TB1 assigned part of Fire Area TB1 85-125-0-38-F Evaluation of Penetrations 1-FP-R1-4-001 and RB1-01j RBI-1 2-FP-R2-4-001 RB2-01j RB2-1 TB1-12 & TB1-13 TB1 TB2-12 & TB2-13 TB1 85-125-0-42-F Evaluation of Control Building Penetration CB-22 and CB-19 CB-23E 0-FP-CB-2-257 TB2-12 & TB2-13 TB1 85-125-0-45-F Sealing Requirements for Penetration 0-FP-CB-2-277 CB-13a CB-2 CB-23 CB-23E 85-125-0-48-F Evaluation of Penetration Seal R2-1-012 RB2-01b RB2-1 RW1-01 a RW-1 PT TB1 85-125-0-53-F Evaluation of Two Conduits in the Unit 2 Reactor RB2-6 RB2-6 Building ECCS Room RB2-01g (N/C) & RB2-01g (N/E) RB2-1 OFP-0021 Downgrade of Rattle Space Wall-to-Sleeve Link-Seals PT TB1 to Non-Fire Rated Status TB2-01 a TB1 RB1-01b, RB1-01a, RB1-01d RB1-1 RB2-01g (N/W), RB2-01b RB2-1 CB-02a, CB-02b, CB-013a, CB-013b CB-2 CB-01a, CB-01b, CB-012a, CB-012b CB-1 OFP-0026 Battery Room Penetration Seals CB-07 CB-7 CB-08 CB-8 CB-05 CB-23E CB-09 CB-9 CB-10 CB-10 704U-M-26 Evaluation of the Use of Nelson CLK and RSW for 3 AOG-1 AOG-1 Hour Fire Barrier Penetration Seals for Spare Open Conduits 84-0622 Penetration Evaluation - SWIS East Wall SWI-1 SWI-1 85-125-0-08-F Diesel Generator Bldg. Pyrocreted Pull Box Enclosures DG-04 DG-4 DG-05 DG-5 85-125-0-14-F PVC Pipe Penetration SW-3-031 in Service Water SW1-1 SW1-1 Intake Structure OUTDOORS BSEP LAR Rev 2 Page A-83

CP&L Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Desiqn Elements Table from Section 3.11.4 Evaluation Title Fire Zone Fire Area 85-125-0-21-F Evaluation of Seals CB-1 -262, 263, 264, 265, 270, 271, CB-06 CB-23E 272, & 273 TB1-01a, TB1-01b, TB2-01a and TB2-01b TB1 85-125-0-23-F Turbine Bldg-Combination Link Seal and Additional TB1-01d & TB2-01d TB1 Moisture Seals No Fire Zone/Area assigned to rattle space 85-125-0-31-F Unit 1 &2 Turbine Bldg/Reactor Wall Thru Pipe Link TB1-01d & TB2-01d TB1 Seals RB1-01a RB1-1 RB1-01b RB1-1 RB2-02a RB2-1 RB2-02b RB2-1 85-125-0-49-F Reactor Bldg.- Eccentric Link-Seal Design RB2-01a RB2-1 No Fire Zone/Area assigned to rattle space 85-125-0-51-F Existing Link Seal Evaluations RB1-01b RB1-1 RB2-01 b RB2-1 RB1-01a RB1-1 RB1-01h RB1-1 89-0149 Evaluation of Service Water Building Penetration Seal SWI-1 SWI-1 SW-3-031 95-00642 Alternative Repair To Fire Barrier Penetration Seals CB-05 CB-23E CB-06 CB-23E No Fire Zone/Area assigned to rattle space 95-01461 Evaluate Fire Rating of Penetrations R2-1-009, T2 RB2-1 b RB2-1 002 and R2-1-018 CB-02a, CB-02b, CB-13a, CB-13b CB-2 TB1-01d TB1 RB2-1a RB2-1 PT TB1 98-00054 Eval. Of Silicone Foam Fire Seals Containing Copper CB-23 CB-23E Pipe CONTROL BUILDING ROOF, No fire zone assigned OUTDOORS 704U-M-22/S1 Evaluation of Conduit Penetrations in Reactor Buildings CB-la, CB-lb, CB-12a, CB-12b CB-1 and Control Building CB-2a, CB-2b, CB-13a, CB-13b CB-2 RB1-01g(S/W & S/C), RB1-01 h(S/W), RB1-10(S) RB1-1 RB2-01 g(N/W & N/C), RB2-01 h(N/W & N/C) RB2-1 85-125-0-41-F Reactor Bldg.-RHR Rooms Penetration Seals RB1-01e & RB1-01f RB1-1 RB2-01e & RB2-01f RB2-1 704U-M-31 Sealing Requirements for Penetration AO-2-057 in AOG AOG-1 AOG-1 Bldg BSEP LAR Rev 2 Page A-84

CP&L Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Table from Section 3.11.4 Evaluation Title Fire Zone Fire Area 85-125-0-02-F U/1 & U/2 Reactor Building ECCS Room Minimum RB1-6 RB1-1 Embedment of Hilti Kwik Bolts for Boot Seal Penetration RB1-1 RB1-1 Fire Seals 85-125-0-05-F Control Building Pull/Junction Box Fire Stop Applications TB1-01a & TB2-01a TB1 All CB-23E Fire Zones CB-23E 85-125-0-10-F Diesel Gen. Bldg. Pyrocrete Enclosure Barriers of Pipe All DG Fire Zones All DG Fire

              & Conduit                                                                                                       Areas 85-125-0-11-F  Diesel Generator Building Evaluation of Thermo-Lag         DG-06                                                DG-6 Installation                                              DG-16                                                DG-16E DG-04                                                DG-4 DG-03                                                DG-3 85-125-0-22-F  Deviation to Design "C" of Specification # 118-003        CB-04                                                 CB-23E CB-06                                                 CB-23E 704U-S-03       12" Diameter Grouted Sleeved Opening, Fire Seal           Generic, All Fire Zones                              All Fire Areas Evaluation 85-125-0-04-F  Cellular Concrete Floor & Wall Blockout Electrical         Generic, All Fire Zones                              All Fire Areas Penetrations Page A-85 BSEP LAR   Rev 2 LAR Rev  2                                                                                                                   Page A-85

Enclosure 5 Revised NFPA 805 Transition Report, Attachment B, NEI 04-02 Table B-2, Nuclear Safety CapabilityAssessment Methodology Review

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review 98 Pages Attached Page B-I BSEP BSEP LAR Rev 22 LAR Rev Page B-1

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3 Deterministic This section discusses a generic deterministic methodology and criteria that licensees can use to Methodology perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific analysis approved by NRC is reflected in the plant's licensing basis. The methodology described in this section is also an acceptable method of performing a post-fire safe shutdown analysis. This methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships. This guidance document, however, does not require the use of a computer database oriented approach. The requirements of Appendix R Sections III.G.1, III.G.2 and III.G.3 apply to equipment and cables required for achieving and maintaining safe shutdown in any fire area. Although equipment and cables for fire detection and suppression systems, communications systems and 8-hour emergency lighting systems are important features, this guidance document does not address them. Additional information is provided in Appendix B to this document. ApPlicability Comments Aliqnment Alignment Basis Statement Aligns Brunswick Steam Electric Plant's (BSEP) Safe Shutdown Methodology was reviewed against the requirements of Appendix R Sections IlI.G, Ill.J, and lII.L as required by 10CFR50.48(b). NRC review and approval of the BSEP safe shutdown methodology is contained in a series of Safety Evaluation Reports. For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with the transition process, NEI 00-01, Revision 1 was one of the references used in developing the circuit analysis procedure, which is now captured in FIR-NGGC-01 01, Revision 2. Except as noted in this document, the plant's methodology meets the guidelines of NEI 00-01. Section 4.0 of BNP-E-9.004 describes the methodology used in the safe shutdown analysis of BSEP. Definitions used in the analysis are presented in Section 3.B including the technical and/or regulatory bases as required. Section 3.B also presents the assumptions and scenarios used in the safe shutdown systems analysis. The safe shutdown performance goals are described in Section 3.C. The Appendix R safety functions identified for BSEP are then described in Section 3.D. Section 2.0 of BNP-E-9.010 describes the methodology used in the safe shutdown analysis of BSEP. BSEP LAR Rev 2 Page B-2

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Definitions used in the analysis are presented in Section 1.3 including the technical and/or regulatory bases as required. Section 1.4 also presents the assumptions and scenarios used in the safe shutdown systems analysis. The safe shutdown performance goals are described in Section 1.5. The Appendix R safety functions identified for BSEP are then described in Section 2.1.1. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.8, 3.C, 3.D and 4.0 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Sections 1.3, 1.4, 2.0 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.0 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1 [A, Intro] Safe This section discusses the identification of systems available and necessary to perform the required Shutdown Systems safe shutdown functions. It also provides information on the process for combining these systems and Path into safe shutdown paths. Appendix R Section IIl.G.l.a requires that the capability to achieve and Development maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" was further clarified in NFPA 805. Appendix R Section IIl.G.l.b requires that repairs to systems and equipment necessary to achieve and maintain cold shutdown be completed within 72 hours. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking. [Refer to hard copy of NEI 00-01 for Figure 3-1] Applicability Comments Alignment Alignment Basis Statement Aligns For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with the transition process, NEI 00-01, Revision 1 was one of the references used in developing the circuit analysis procedure, which is now captured in FIR-NGGC-0101, revision 2. Except as noted in this document, the plant's methodology meets the guidelines of NEI 00-01. As pointed out in Section 4.2.1.2, given a fire, NFPA 805 does not require a plant to transition to cold shutdown. The fire area-by-fire area assessment documents the method of accomplishment of the NFPA 805 performance goals (including the transition to cold shutdown). During transition, Bunswick BSEP LAR Rev 2 Page B-3

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review did not attempt to change the safe shutdown analysis to remove equipment/cables (and compliance strategies) that were only required to achieve and maintain cold shutdown. However, as allowed by the NFPA 805 change process and the revised license condition, Brunswick may revise these strategies post-transition. Note that these cold shutdown actions are not considered VFDRs since they are not required to achieve safe and stable conditions. Safe and stable is considered to have been achieved when the reactor is shutdown (keff < 0.99), a method of RPV inventory control has been established, and the suppression pool cooling mode of RHR has been established. In Section 4.0 of BNP-E-9.004, the safe shutdown functions described in Section 3.D establish the framework for identifying those systems and components necessary for safe shutdown. This section describes the process used to identify these systems and components. The principal steps in this process are: (1) Relate systems to the safe shutdown functions. (2) Identify those components in each system required for its successful achievement of the safe shutdown function. (3) Completion of circuit analysis for required components to identify necessary cables for local or normal operating stations. In Section 2.0 of BNP-E-9.010, the safe shutdown functions described in Section 2.1.1 establish the framework for identifying those systems and components necessary for safe shutdown. This section describes the process used to identify these systems and components. The principal steps in this process are: (1) Relate systems to the safe shutdown functions. (2) Identify those components in each system required for its successful achievement of the safe shutdown function. (3) Completion of circuit analysis for required components to identify necessary cables for local or normal operating stations. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.0 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 2.1.1 and 2.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.0 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1 [B, Goals] Safe The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and BSEP LAR Rev 2 Page B-4

CP&L Attachment B- NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Shutdown Systems components remains free of fire damage for a single fire in any single plant fire area. This goal is and Path accomplished by determining those functions important to achieve and maintain hot shutdown. Safe Development shutdown systems are selected so that the capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following: Reactivity control Pressure control systems Inventory control systems Decay heat removal systems Process monitoring Support systems

                             - Electrical systems
                             - Cooling systems These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment. If these functions are preserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.

Applicability Comments Alignment Alignment Basis Statement Aligns BNP-E-9.004 identifies the safe shutdown performance goals and describes the safe shutdown functions. BNP-E-9.010 identifies the safe shutdown performance goals and describes the safe shutdown functions. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.C and 3.D BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sectioins 2.1.1 and 2.2 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteda shall be included. Availability and reliability of equipment selected shall be evaluated. BSEP LAR Rev 2 Page B-5

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.1 [C, Spurious In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated Operations] Safe circuits with the potential for spurious equipment operation and/or loss of power source, and the Shutdown Systems common enclosure failures. Spurious operations/actuations can affect the accomplishment of the and Path post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious Development operations of concern are the following:

                             - A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability
                             - A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.

Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment. Common power source and common enclosure concerns could also affect these and must be addressed. ApDlicability Comments Alignment Alignment Basis Statement Aligns BSEP has considered spurious operation, common power sources, and common enclosure concerns that would cause a circuit to be considered an associated circuit. During the re-validation, the high-low interface definition from the previous SSA was carried forward for conservatism. Thus, some components are classified as high-low interfaces which do not meet the above definition since their spurious opening will not result in a rupture of downstream piping and a subsequent intersystem LOCA. Brunswick may choose to remove the classification of these components as high-low interfaces at a future date. RCS isolation valves (such as the suction valves) are defined as high/low pressure interface boundary valves if their spurious operation could lead to the rupture of low pressure piping or a loss of RCS inventory that exceeds the RCS makeup capability. Such interface boundary valves are subject to more stringent circuit analysis criteria, and are identified in FSSPMD by the HLP flag. This high/low pressure interface boundary valve definition is conservative with respect to that in in Appendix C of NEI 00-01 and NFPA-805 FAQ 06-0006. BNP-E-9.010 addresses the implications of spurious operations in the identification of safe shutdown components, cables, and circuits. BNP-E-9.010 identifies the systems and components necessary to achieve and maintain safe shutdown. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 4.B and 4.C, Section 3.E BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 3.2.3, Section 2.2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.4 and 3.34 (NSCA) BSEP LAR Rev 2 Page B-6

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1 Criteria / The following criteria and assumptions may be considered when identifying systems available and Assumptions necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths. AIDlicabilitv Comments Alignment Statement Aliqnment Basis N/A This is generic introductory information and contains no specific requirements. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.1 [GE BWR [BWR] GE Report GE-NE-T43-00002-00-01-RO1 entitled "Original Safe Shutdown Paths For The Paths] BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section Ill.G.1 of 10CFR 50, Appendix R. Any of the shutdown paths (methods) described in this report are considered to be acceptable methods for achieving redundant safe shutdown. Applicability Comments Alignment BSEP LAR Rev 2 Page B-7

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Statement Alignment Basis Aligns For each safe shutdown system, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. This identification included those branch flow paths which must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. A functional block diagram was produced for each function showing the potential success paths for achieving the required function; and illustrate the redundant combinations of safe shutdown equipment providing the multiple process paths which fulfill the BSEP safe shutdown functions. For each safe shutdown system, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. This identification included those branch flow paths which must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. These flow paths are reflected in the CAFTA fault tree used with the ARC software to develop the NSCA. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.B BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.1.2 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.10 [ManualI Manual initiation from the main control room or emergency control stations of systems required to Automatic Initiation of achieve and maintain safe shutdown is acceptable where permitted by current regulations or Systems] approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option. Applicability Comments Alignment Alignment Basis Statement BSEP LAR Rev 2 Page B-8

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Aligns Manual initiation of equipment and systems is credited in the safe shutdown analysis. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (18, 23) (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.11 [Multiple Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and Affected Units] maintain safe shutdown for each affected unit must be demonstrated. Applicability Comments Aligqnment Alignment Basis Statement Aligns Since BSEP is a two unit site with shared systems, the plant's SSA and safe shutdown procedures address shutdown of both units for each analyzed fire / analysis area. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.004, Safe Shutdown Analysis Report BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. BSEP LAR Rev 2 Page B-9

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.2 [SRVs / LP [BWR] GE Report GE-NE-T43-00002-00-03-RO1 provides a discussion on the BWR Owners' Group Systems] (BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of 10CFR50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000. ADplicability Comments Alignment Alignment Basis Statement Aligns BNP-E-9.004 describes SRV operation with Core Spray, ADS, and RHR in the LPCI mode. BNP-E-9.010 describes SRV operation with Core Spray, ADS, and RHR in the LPCI mode. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.E.3, 3.E.4 and 3.E.5 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Sections 2.2.3.3, 2.2.3.4, and 2.2.3.5 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.3 [Pressurizer [PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be Heaters] maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled. Aoplicability Comments Alignment Alignment Basis Statement N/A BSEP is a BWR. This guidance is specific to PWRs. Comments BSEP LAR Rev 2 Page B- 10

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.4 [Alternative The classification of shutdown capability as alternative shutdown is made independent of the Shutdown Capability] selection of systems used for shutdown. Alternative shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections III.G.1 and III.G.2 may be supplemented by the use of manual actions to the extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations, as appropriate. These may also be used in conjunction with alternative shutdown capability. Applicability Comments Alignment Alignment Basis Statement Aligns BNP-E-9.004 describes BSEPs alternative shutdown methodology. BNP-E-9.010 describes BSEPs alternative shutdown methodology. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.G BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.2.3.29 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. BSEP LAR Rev 2 Page B-11I

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.5 [Initial At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) Conditions] are assumed operable and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress. The units are assumed to be operating at full power under normal conditions and normal lineups. Applicability Comments Aligqnment Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to BSEP. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1(2) (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.6 [Other Events No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, in Conjunction with earthquake), single failures or non-fire induced transients need be considered in conjunction with the Fire] fire. Applicability Comments Alignment Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to BSEP. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (6,7,8) (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-12

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.7 [ Offsite For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire Power] damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours. ApDlicability Comments Alignment Statement Alignment Basis Statement Aligns For fire areas that use redundant shutdown capabilities offsite power is credited unless the fire impacts equipment required to support offsite power. If the fire impacts offsite power, at least one onsite power source is available to provide the required power. For areas that use alternative shutdown, a LOOP is assumed. In the analysis the LOOP is not credited for preventing or terminating spurious operations or positioning SSE in its required position. Steps in the procedures insure that the appropriate actions are taken to line up SSE and deal with potential spurious equipment operations. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 1.1.2, 1.4.1, and 1.5.2 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear BSEP LAR Rev 2 Page B-13

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.8 [Safety- Post-fire safe shutdown systems and components are not required to be safety-related. Related Equipment] Applicability Comments Alignment Statement Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to BSEP. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.9 [72 Hour The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor Coping] scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can be repaired or made operable within 72 hours using onsite capability to achieve cold shutdown. Applicability Comments Alignment Statement Alignment Basis Statement Aligns with Intent NFPA 805 does not require a plant to transition to cold shutdown in the event of a fire. The fire area-by-fire area assessment documents the method of accomplishment of the NFPA 805 performance goals, including an optional transition to cold shutdown. For all fires at BSEP, the systems and equipment required to place the plant in a safe and stable condition are available following a fire occurring while the plant is at power without regard to a specific mission time or event coping duration. Comments BSEP LAR Rev 2 Page B-14

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 5.0 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 1.5.2 and 2.2.3.29 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2 Shutdown The following discussion on each of these shutdown functions provides guidance for selecting the Functions systems and equipment required for safe shutdown. For additional information on BWR system selection, refer to GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths for the BWR." Applicabilitv Comments Alignment SAtegment Alignment Basis Statement Aligns This is an introductory section with no specific requirements. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.1 Reactivity [BWR] Control Rod Drive System Control The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor. [PWR] Makeup/Charging BSEP LAR Rev 2 Page 13-15

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review There must be a method for ensuring that adequate shutdown margin is maintained by ensuring borated water is utilized for RCS makeup/charging. Applicability Comments Alignment Alignment Basis Statement Aligns Reactivity control for safe shutdown and alternative shutdown credits the ability to scram the reactor. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.D.1 and 3.G.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.2.1.1 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.2 Pressure The systems discussed in this section are examples of systems that can be used for pressure Control Systems control. This does not restrict the use of other systems for this purpose. [BWR] Safety Relief Valves (SRVs) The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are operated manually. Automatic initiation of the Automatic Depressurization System is not a required function. [PWR] Makeup/Charging RCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiating charging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is acceptable. Applicability Comments Alignment Alignment Basis Statement Aligns SRV operations associated with altemate shutdown cooling, core spray system, ADS, and residual BSEP LAR Rev 2 Page B-16

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review heat removal are addressed. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.D.1 and 3.G.1 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.2.1.1 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.3 Inventory [BWR] Systems selected for the inventory control function should be capable of supplying sufficient Control reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required. [PWR]: Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown. Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable. Automatic initiation functions are not required. Applicability Comments BSEP LAR Rev 2 Page 13-17

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.4 Decay Heat [BWR] Systems selected for the decay heat removal function(s) should be capable of: Removal

                             - Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure.
                             - Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool).
                             - Removing sufficient decay heat from the reactor to achieve cold shutdown.

[PWR] Systems selected for the decay heat removal function(s) should be capable of:

                             - Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this entails utilizing natural circulation in lieu of forced circulation via the reactor coolant pumps and controlling steam release via the Atmospheric Dump valves.
                             - Removing sufficient decay heat from the reactor to reach cold shutdown conditions.

This does not restrict the use of other systems. ADplicabilitv Comments Alignment Alignment Basis Statement Aligns Systems used for decay heat removal functions through cold shutdown conditions are addressed. HPCI, RCIC, LPCI, and safety/relief valves provide the capability to restore and maintain reactor vessel level and control pressure. The RHR system removes decay heat from the torus and provides a means for removing decay heat, maintain reactor coolant temperatures below 212 *F, and provide reactor coolant makeup water. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.C, D.2, D.4. D.5, and E.5 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 2.2.3.1 - 2.2.3.5 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection BSEP LAR Rev 2 Page B- 18

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.5 Process The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1, Monitoring Section IX "Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessary to perform and control the functions specified in Appendix R Section IIIL.. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures (including Abnormal Operating Procedures). [BWR]

                    - Reactor coolant level and pressure
                    - Suppression pool level and temperature
                    - Emergency or isolation condenser level
                    - Diagnostic instrumentation for safe shutdown systems
                    - Level indication for tanks needed for safe shutdown

[PWR]

                    - Reactor coolant temperature (hot leg / cold leg)
                    - Pressurizer pressure and level
                    - Neutron flux monitoring (source range)
                    - Level indication for tanks needed for safe shutdown
                    - Steam generator level and pressure
                    - Diagnostic instrumentation for safe shutdown systems The specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. prescriptive), and systems and paths selected for safe shutdown.

Applicability Comments Alignment Alignment Basis Statement Aligns In order to achieve and maintain safe shutdown conditions, the operator must be able to monitor various plant parameters. These parameters provide the information required by the operators in order to perform required system transitions and essential operator actions. This function ensures the instrumentation required to monitor reactor level, reactor pressure, suppression pool level, and suppression pool temperature is available following any fire. Process monitoring instrumentation and local RHR instrumentation provide adequate monitoring during the cooldown to cold shutdown. Comments BSEP LAR Rev 2 Page B- 19

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D.6, E.5 and E.6 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.2.3.6 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6 Support [Blank Heading - No specific guidance] Systems ADnlicability Comments Alignment Alignment Basis Statement N/A Support system requirements are addressed under the corresponding NEI 00-01 sub-section. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.1 Electrical AC Distribution System Systems Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busses either directly from the busses or through step down transformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sources or the emergency diesel generator depending on BSEP LAR Rev 2 Page B-20

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a loss of offsite power. Refer to Section 3.1.1.7. DC Distribution System Typically, the 125VDC distribution system supplies DC control power to various 125VDC control panels including switchgear breaker controls. The 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters. These distribution panels typically supply power for instrumentation necessary to complete the process monitoring functions. For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the diesel generators to become operational. Once the diesels are operational, the 125 VDC distribution system can be powered from the diesels through the battery chargers. [BWR] Certain plants are also designed with a 250VDC Distribution System that supplies power to Reactor Core Isolation Cooling and/or High Pressure Coolant Injection equipment. The DC control centers may also supply power to various small horsepower Appendix R safe shutdown system valves and pumps. Ifthe DC system is relied upon to support safe shutdown without battery chargers being available, it must be verified that sufficient battery capacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required to operate). Applicability Comments Alignment Alignment Basis Statement Aligns The Electrical Distribution System provides 4160VAC, 480VAC, 120VAC and 250VDC/125VDC power from off-site (115KV Grid) and onsite sources (EDGs) to safe shutdown electrical loads. The fuel oil systems associated with the onsite power supplies are also included in the analysis. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D.7, E.8 and E.9 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 2.2.3.8 - 2.2.3.10 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 Cooling HVAC Systems Systems [HVAC] BSEP LAR Rev 2 Page B-21

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer's literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents). HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations. Typical uses include:

                             - Main control room, cable spreading room, relay room
                             - ECCS pump compartments
                             - Diesel generator rooms
                             - Switchgear rooms Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.

Applicability Comments Alignment Alignment Basis Statement Aligns The -17 ft. elevation (east side) of the Reactor Building and the diesel generator cells are cooled with the normal HVAC equipment, or at least some portion of it. The emergency cooling system maintains the areas containing ECCS pumping equipment within the required temperature range at all times, including during postulated DBA conditions. The Control Building HVAC Systems are designed to permit continuous occupancy of the Control Room, computer rooms, and the electronic workrooms under normal plant operating conditions and under postulated DBA conditions. Specific Systems are listed in the Safe Shutdown Analysis. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section B.1 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.2.3.11 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 Cooling Various cooling water systems may be required to support safe shutdown system operation, based BSEP LAR Rev 2 Page B-22

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Systems [Main on plant-specific considerations. Typical uses include: Section] - RHR/SDC/DH Heat Exchanger cooling water

                             - Safe shutdown pump cooling (seal coolers, oil coolers)
                             - Diesel generator cooling
                             - HVAC system cooling water Applicability             Comments Alignment Statement                 Alignment Basis Statement Aligns                    The Service Water (SW) System provides essential cooling water for operation of the diesel generators and for cooling of the RHR pumps, area coolers, and heat exchangers.

Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D.7 and E.7 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.2.3.7 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.3 Methodology for Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown Shutdown System systems and developing the shutdown paths. Selection The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis: [Refer to hard copy of NEI 00-01 for Figure 3-2] Applicability Comments Alignment Statement Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-23

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.1 Identify safe Review available documentation to obtain an understanding of the available plant systems and the shutdown functions functions required to achieve and maintain safe shutdown. Documents such as the following may be reviewed:

                             - Operating Procedures (Normal, Emergency, Abnormal)
                             - System descriptions
                             - Fire Hazard Analysis
                             - Single-line electrical diagrams
                             -Piping and Instrumentation Diagrams (P&IDs)

[BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR" Applicability Comments Alignment Alignment Basis Statement Aligns Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire Hazard Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system comprising the safe shutdown paths, the mechanical or electrical equipment required for the operation of the system and the equipment whose spurious operation could affect the performance of the safe shutdown systems were identified. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.D BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.2.1 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components BSEP LAR Rev 2 Page B-24

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.2 Identify Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of Combinations of systems capable of achieving the safe shutdown functions of reactivity control, pressure control, Systems that Satisfy inventory control, decay heat removal, process monitoring, and support systems such as electrical Each Safe Shutdown and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other Function systems. In addition to achieving the required safe shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function. ADnlicabilitv Comments Alignment Alignment Basis Statement Aligns Each function and its relationship to the safe shutdown performance goals is identified. The systems and components required to attain safe shutdown in case of fire are described. This analysis identified the potential spurious operation candidates and placed them into one of the following two categories: (1) Spurious operation candidates which could affect proper safe shutdown system operation; and (2) Spurious operation candidates which could cause an uncontrolled loss of primary coolant. Those spurious candidates which fall into the first category were addressed by including these devices on the safe shutdown equipment list for the affected safe shutdown system and analyzing them as a safe shutdown component. Those spurious candidates which fall into the second category were analyzed on a case-by-case basis. For each system identified, plant P&lDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. This identification included those branch flow paths which must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. From this information, a list was compiled which identified all components required for each system's performance of its safe shutdown function. For each component, the following information was identified: (1) System (2) Train (3) Mode of Safe Shutdown (4) Required Position (5) Category Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D, E, F.2 and 4.B BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 2.1.2, 2.2.1, and 2.2.3 BSEP LAR Rev 2 Page B-25

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.3 Define Select combinations of systems with the capability of performing all of the required safe shutdown Combinations of functions and designate this set of systems as a safe shutdown path. In many cases, safe shutdown Systems for Each paths may be defined on a divisional basis since the availability of electrical power and other support Safe Shutdown Path systems must be demonstrated for each path. ApDlicabilit Comments Alignment Alignment Basis Statement Aligns Each function and its relationship to the safe shutdown performance goals is identified. The systems and components required to attain safe shutdown in case of fire are described. For each system identified, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. This identification included those branch flow paths which must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. From this information, a list was compiled which identified all components required for each system's performance of its safe shutdown function. Fault trees which are part of the CAFTA logic provide a graphical display of how the components of the safe shutown systems are aligned to meet the performance goals. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D, E and 4.6 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.2.3 FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-26

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.4 Assign Assign a path designation to each combination of systems. The path will serve to document the Shutdown Paths to combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to Each Combination of this document (NEI 00-01) for an example of a table illustrating how to document the various Systems combinations of systems for selected shutdown paths. Applicability Comments Alignment Alignment Basis Statement Aligns with intent Each function and its relationship to the safe shutdown performance goals is identified. The systems and components required to attain safe shutdown in case of fire are described. For each system identified, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. This identification included those branch flow paths which must be isolated and remain isolated to assure that flow will not be diverted from the primary flow path. From this information, a list was compiled which identified all components required for each system's performance of its safe shutdown function. Fault trees which are part of the CAFTA logic provide a graphical display of how the components of the safe shutown systems are aligned to meet the performance goals. Each combination of systems or success paths is not assigned a path designation. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.004, Safe Shutdown Analysis Report Sections 3.D, E and 4.B BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 2.2.2 and 2.2.3 FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection BSEP LAR Rev 2 Page B-27

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result inrthe maloperation of those components needed to meet the nuclear safety criteda shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2 Safe Shutdown The previous section described the methodology for selecting the systems and paths necessary to Equipment Selection achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix R function. The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the maloperation of the safe shutdown systems. APPlicabilitY Comments Aliqnment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nucdear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1 Criteria / Consider the following criteria and assumptions when identifying equipment necessary to perform the Assumptions required safe shutdown functions: AplDlicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments BSEP LAR Rev 2 Page B-28

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.1 [Primary 3.2.1.1 Safe shutdown equipment can be divided into two categories. Equipment may be categorized Secondary as (1) primary components or (2) secondary components. Typically, the following types of Components] equipment are considered to be primary components:

                             - Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.
                             - All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)
                             - Power supplies or other electrical components that support operation of primary components (i.e.,

diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.). Secondary components are typically items found within the circuitry for a primary component. These provide a supporting role to the overall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either an interiock or input signal processor. Examples of secondary components include flow switches, pressure switches, temperature switches, level switches, temperature elements, speed elements, transmitters, converters, controllers, transducers, signal conditioners, hand switches, relays, fuses and various instrumentation devices. Determine which equipment should be included on the Safe Shutdown Equipment List (SSEL). As an option, include secondary components with a primary component(s) that would be affected by fire damage to the secondary component. By doing this, the SSEL can be kept to a manageable size and the equipment included on the SSEL can be readily related to required post-fire safe shutdown systems and functions. Applicability Comments Alignment Aliqnment Basis Statement Aligns Components are not identified as primary or secondary. Components providing a "secondary" function are either identified as safe shutdown components and linked to the "primary" component in the fault tree, or have their applicable cables assigned to the primary component. Within FSSPMD, such "secondary" components are idenitied as associated circuits for the applicable "primary" component. Cable failures affecting these associated circuits "cascade" via the fault tree to ensure all potential affects on the "primary" component are captured. BSEP LAR Rev 2 Page B-29

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review The detailed structure and format of the Safe Shutdown Equipment List has been defined in the Fire Safe Shutdown Program Manager Database and is described in the FSSPMD User Manual. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.2 [Fire Damage 3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact to Mechanical their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping Components (not materials, including tubing with brazed or soldered joints, are not included in this assumption). Fire electrically damage should be evaluated with respect to the ability to manually open or close the valve should supervised)] this be necessary as a part of the post-fire safe shutdown scenario. Applicability Comments Alignment Statement Alignment Basis Aligns Due to the substantial nature of equipment and the nature and location of combustibles, fire will not not impact the pressure boundary function. A fire does not cause a manual valve to change its position. Manual stroking of a valve once the fire is extinguished is evaluated as part of the manual action feasibiltiy study. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.1.3 and 9.4.1 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-GAP-001 Closed 06-23-12 El BSEP LAR Rev 2 Page B-30

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review VFDR Disposition For any rising stem valves that require post fire manual Change Package BNP-0240 evaluated rising stem operation within the fire area of concern, ensure the valves and determined that there are no risking stem feasibility criteria take into account the increased valves listed in OASSD-00, Revision 41, Section 5.5.5 friction due to the lubricant being burned away. This that require manual operation after being in the fire will be incorporated into an upcoming revision to BNP- area. An NCR has been generated to incorporate this E-9.007 (Section 3.2.1.2). into a revision of BNP-E-9.007. FRE/Chanae Eval/Mod Corrective Action NCR 482987

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.3 [Manual Valve Assume that manual valves are in their normal position as shown on P&lDs or in the plant operating Positions] procedures. Applicability Comments Aliqnment Statement Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to BSEP. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear BSEP LAR Rev 2 Page B-31

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.4 [Check Assume that a check valve closes in the direction of potential flow diversion and seats properly with Valves] sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions. ApDlicabilitv Comments Aliqnment Aliqnment Basis Statement Aligns FIR-NGGC-001 identifies that properly oriented check valves credited as system boundaries should be included in the SSEL, and that those in the flow path need not be included. Check Valves credited as boundaries are included in the SSEL, but the assumption that they are leak tight is inherent in the analysis. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.1.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.5 [Instrument Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow Failures] transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit. Applicability Comments Aliqnment Alignment Basis Statement Aligns Instruments exposed to the fire are assumed to fail. It is a generally accepted practice (that can be verified based on a review of the fire area analysis) that instruments are assumed to fail to their worst case position. Comments BSEP LAR Rev 2 Page B-32

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 1.4.6.1 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.2 and 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.6 [Spurious Identify equipment that could spuriously operate or mal-operate and impact the performance of Components] equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process. Aonlicabilitv Comments Alignment Statement Alignment Basis Statement Aligns Electrically operated or controlled valves or dampers in the flow paths whose spurious operation could adversely affect system operation are included on the SSEL. This analysis identified the potential spurious operation candidates and placed them into one of the following two categories: (1) Spurious operation candidates which could affect proper safe shutdown system operation; and (2) Spurious operation candidates which could cause an uncontrolled loss of primary coolant. Those spurious candidates which fall into the first category were addressed by including these devices on the safe shutdown equipment list for the affected safe shutdown system and analyzing them as a safe shutdown component. Those spurious candidates which fall into the second category were analyzed on a case-by-case basis. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.2 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.1.2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.1.3 and 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-33

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.7 [Instrument Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a Tubing] result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area. Applicability Comments Alignment Alignment Basis Statement Aligns FIR-NGGC-0101 provides direction for evaluating the fire effects on instrument tubing. FSSPMD documents tubing routing to ensure the impact of this issue is evaluated. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.7 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.2 Methodology for Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown Equipment Selection equipment. Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis: [Refer to hard copy of NEI 00-01 for Figure 3-3] BSEP LAR Rev 2 Page B-34

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.1 Identify the Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each System Flow Path for shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept. Each Shutdown Path Applicability Comments Alignment SAtegment Alignment Basis Statement Aligns For each system identified as necessary to perform a SSD function, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.8 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 1.6.4.8, 2.1.2, and 2.2.3 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.5 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection BSEP LAR Rev 2 Page B-35

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.2 Identify the Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to Equipment in Each assure that all equipment in each system's flow path has been identified. Assure that any equipment Safe Shutdown that could spuriously operate and adversely affect the desired system function(s) is also identified. If System Flow Path additional systems are identified which are necessary for the operation of the safe shutdown system Including Equipment under review, include these as systems required for safe shutdown. Designate these new systems That May Spuriously with the same safe shutdown path as the primary safe shutdown system under review (Refer to Operate and Affect Figure 3-1). System Operation Applicability Comments Alignment Statement Alignment Basis Aligns For each system identified as necessary to perform a SSD function, plant P&IDs, system descriptions, and one-line diagrams were used to identify the safe shutdown flow paths and operational characteristics that must be established to accomplish the desired safe shutdown functions. During the analysis of the safe shutdown flow paths, those components whose spurious operation could impair safe shutdown system operability were also identified. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.B BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sections 1.6.4.8, 2.1.2, and 2.2.3 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteda shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.3 Develop a List Prepare a table listing the equipment identified for each system and the shutdown path that it of Safe Shutdown supports. Identify any valves or other equipment that could spuriously operate and impact the BSEP LAR Rev 2 Page B-36

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Equipment and operation of that safe shutdown system. Assign the safe shutdown path for the affected system to Assign the this equipment. During the cable selection phase, identify additional equipment required to support Corresponding the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this System and Safe additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an Shutdown Path(s) example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe Designation to Each. shutdown and it documents various equipment-related attributes used in the analysis. Applicability Comments Alignment Alignment Basis Statement Aligns Information in FSSPMD includes the component's power supply, fire zone location, normal and required positions, required cables, and associated circuits. A SSEL was compiled which identified all components required for each system's performance of its safe shutdown function. For each component, the following information was identified: (1) System (2) Train (3) Mode of Safe Shutdown (4) Required Position (5) Category The detailed structure and format of the Safe Shutdown Equipment List has been defined in the Fire Safe Shutdown Program Manager Database and is described in the FSSPMD User Manual. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Sections 4.B BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 2.1.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 9.1.4 and 9.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.4 Identify Collect additional equipment-related information necessary for performing the post-fire safe Equipment shutdown analysis for the equipment. In order to facilitate the analysis, tabulate this data for each BSEP LAR Rev 2 Page B-37

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Information Required piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a for the Safe SSEL. Examples of related equipment data should include the equipment type, equipment Shutdown Analysis description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as the following may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, and spurious operation concern. Applicability Comments Alignment Alignment Basis Statement Aligns The information identified as needed for performing safe shutdown analysis on the component identified on the SSEL is contained in FSSPMD. This can be verified on a component basis through reports that can be generated through FSSPMD. Information in FSSPMD included the component's power supply, fire zone location, normal and required positions, required cables, and associated circuits. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.B BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 2.1.2 FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated. NEI 00-01 Ret NEI 00-01 Guidance 3.2.2.5 Identify In the process of defining equipment and cables for safe shutdown, identify additional supporting Dependencies equipment such as electrical power and interlocked equipment. As an aid in assessing identified Between Equipment, impacts to safe shutdown, consider modeling the dependency between equipment within each safe Supporting shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram Equipment, Safe (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these Shutdown Systems relationships. and Safe Shutdown Paths. Applicability Comments BSEP LAR Rev 2 Page B-38

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Alignment Alignment Basis Statement Aligns Power supplies are identified and documented in FSSPMD. Cables that are associated with a component because of interlocks or permissive requirements are documented in FSSPMD. The safe shutdown logic has been documented in a fault tree logic (FTL) file which models the safe shutdown functions, systems and components. This approach has been utilized in lieu of the traditional approach where system level, component level, and electrical logic diagrams are used to demonstrate a successful safe shutdown path. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3 Safe Shutdown This section provides industry guidance on the recommended methodology and criteria for selecting Cable Selection and safe shutdown cables and determining their potential impact on equipment required for achieving Location and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the maloperation of safe shutdown BSEP LAR Rev 2 Page B-39

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review equipment are identified and that these cables are properly related to the safe shutdown equipment whose functionality they could affect. Through this cable-to-equipment relationship, cables become part of the safe shutdown path assigned to the equipment affected by the cable. Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1 Criteria / To identify an impact to safe shutdown equipment based on cable routing, the equipment must have Assumptions cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment. Consider the following criteria when selecting cables that impact safe shutdown equipment: Applicability Comments Alignment Alignment Basis Statement BSEP LAR Rev 2 Page B-40

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nudlear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and intemal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.1 [Cable The list of cables whose failure could impact the operation of a piece of safe shutdown equipment Selection] includes more than those cables connected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, and equipment status indication cables related to the equipment. Consider reviewing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post-fire safe shutdown. As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4. Applicability Comments Aliqnment Alignment Basis Statement Aligns FIR-NGGC-0101 provides direction for assigning cables to components. This process is documented in BNP-E-9-010 and in FSSPMD. Comments BSEP LAR Rev 2 Page B-41

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Circuit Information Form Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and intemal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.2 [Cables In cases where the failure (including spurious actuations) of a single cable could impact more than Affecting Multiple one piece of safe shutdown equipment, include the cable with each piece of safe shutdown Components] equipment. Applicability Comments Alignment SAtegment Alignment Basis Statement Aligns Circuit analysis is performed independently on individual components, so cables affecting more than one component is identified with each impacted component. Normal and alternate power supplies, and associated circuits are documented in FSSPMD. Comments BSEP LAR Rev 2 Page B-42

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report 4.C FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Circuit Information Form Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.3 [Isolation Electrical devices such as relays, switches and signal resistor units are considered to be acceptable Devices] isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function. Applicability Comments Alignment Alignment Basis Statement Aligns An isolation device is a component in a circuit which prevents malfunctions in one section of an electrical circuit from causing unacceptable effects in other sections of the circuit or other circuits. Acceptable isolation devices for power circuits are single isolation devices actuated by fault currents (breakers and fuses). For low energy control and instrumentation circuits, acceptable isolation devices are those actuated by fault currents (e.g., fuses or circuit breakers), relays, control switches, transducers, isolation amplifiers, current transformers, diodes, and fiber optic couplers. BSEP LAR Rev 2 Page B-43

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 3.0 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.4 [Identify "Not Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., Required" Cables] annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these circuits is necessary. However, they must be isolated from the component's control scheme in such a way that a cable fault would not impact the performance of the circuit. Applicabilitv Comments Alignment Alignment Basis Statement Aligns In FSSPMD cables that are not required for safe shutdown have an "A" or an "NA" entered in the FMEA section of the circuit information form. The "A" indicates the the component "achieves" its safe shutdown function even if the cable is damaged by a fire. The "NA" indicates that the cable is not part os the safe shutdown circuit. Comments BSEP LAR Rev 2 Page B-44

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 and Attachment 1 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.5 [Identification For each circuit requiring power to perform its safe shutdown function, identify the cable supplying of Power Supplies] power to each safe shutdown and/or required interlock component. Initially, identify only the power cables from the immediate upstream power source for these interlocked circuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns. ApDlicability Comments Alignment Alignment Basis Statement Aligns The power cables for individual components are listed in the circuit analysis for that component if BSEP LAR Rev 2 Page B-45

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review power is needed for the component to perform its safe shutdown function. Power supplies are linked to their components in FSSPMD in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" portion of the Circuit Information Form. Standard note "A" is entered for a power supply that is required for the component to perform its safe shutdown function. The power supply requirement is modeled in the fault tree. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 and Attachment 1 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.6 [ESFAS The automatic initiation logics for the credited post-fire safe shutdown systems are not required to Initiation] support safe shutdown. Each system can be controlled manually by operator actuation in the main control room or emergency control station. Ifoperator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function. Applicability Comments BSEP LAR Rev 2 Page B-46

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Alignment Alignment Basis Statement Aligns For components with an Engineered Safeguards Actuation Signal input (ESFAS or ESAS), cable faults occurring between the ESFAS contacts and the ESFAS master relay associated with the ESFAS signal is included in the circuit analysis. However, the analysis will not include the initiating logic circuits and input circuits to the safeguards cabinets. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.7 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.7 [Circuit Cabling for the electrical distribution system is a concern for those breakers that feed associated Coordination] circuits and are not fully coordinated with upstream breakers. With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component. For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also support the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source and not the connected load. Similarly, the cable feeding the MCC from BSEP LAR Rev 2 Page B-47

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review the load center would also be necessary to support the load center. Applicability Comments Alignment Statement Alignment Basis Aligns The guidelines that are used in the evaluation of the common power supplies are as follows:

                              - Using the single-line drawings, ensure that all safe shutdown power supplies required have been included.
                              - For each safe shutdown power supply, review the following documents (as necessary): existing short circuit calculations, load studies, coordination calculations, protective device setting sheets, and time current curves as appropriate to confirm proper coordination between upstream and downstream protective devices to ensure that they are up to date.
                              - In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis shall be considered.
                              - Breaker coordination shall consider both the instantaneous and thermal trip regions of the breaker curves. However, if the analyst must choose between instantaneous and thermal, the instantaneous region of the curve should be the governing consideration.
                              - For cases in which coordination between series protective devices cannot be demonstrated, a common power supply associated circuit are assumed to exist. These circuits are dispositioned by one of the following means:
                             - Demonstrate coordination by refining the available short circuit current and/or trip device characteristics.
                             - Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g., equipment located in same fire area as power supply).
                             - Identify readily achievable protective device setting changes (including changes in fuse size and/or clearing characteristics) that will establish coordination.
                             - Incorporate the Associated Circuits and Cables into the post-fire safe shutdown analysis when protection devices do not provide the desired coordination.
                              - Site specific short circuit and coordination calculations shall be updated as necessary to fully document where coordination is credited for post-fire safe shutdown. The electrical portion of the Safe Shutdown Analysis Report, completed under the Fire Protection Initiatives Project, includes a complete description of the common power supply associated circuits analysis, including reference to applicable supporting calculations and documents.

Comments Reference Document Doc Detail BNP-E-6.085, Unit 2 125/250V DC Coordination/Protection Calculation BNP-E-6.095, Unit 1 125/250V DC Coordination/Protection Calculation BNP-E-8.010, AC Coordination Study BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supportina Detail BSEP LAR Rev 2 Page B-48

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review B2-GAP-002 Closed 04-27-12 VFDR Disposition Section 3.3.1.1.6 is a new section in revision two of NEI The Fire Safety Analyses for SW-1 for Unit 1, 1FP-1 127 00-01, and discusses the use of an exclusion analysis and Unit 2, 2FP-1 197 have been revised to incorporate in lieu of rigorous circuit analysis. For a fire in SW-01, a VFDR for this area in each unit. Using the an exclusionary analysis was done to justify the performance-based approach to resolve the VFRD availability of only one service water pump per unit. eliminates the need to use the exclusion analysis and Post EPU, the thermal-hydraulic analysis requires two revise BNP-E-9.010 before LAR submittal. service water pumps per unit, but the pre-transiton exemption for this area was based on only one pump being required per unit. NCR 482987 concluded that the availabilty of offsite power and other equipment not credited in the thermal-hydraulic calculation, only one service water pump would be required. This exclusion analysis needs to be formally documented and reflected in the compliance assessment summary for Fire Area SW-01. This should be done in the next revision of BNP-E-9.010, and before the LAR. FREBChan2e EvalrMod Se Corrective Action Ci 88191 NCR 482987

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance BSEP LAR Rev 2 Page B-49

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 3.3.2 Associated Appendix R, Section III.G.2, requires that separation features be provided for equipment and cables, Circuit Cables including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve hot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 6.1.6. They are as follows:

                             - Spurious actuations
                             - Common power source
                             - Common enclosure.

Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [A] Associated Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose Circuit Cables - failure could cause the spurious actuation/mal-operation of equipment whose operation could affect Cables Whose safe shutdown. These cables are identified in Section 3.3.3 together with the remaining safe Failure May Cause shutdown cables required to support control and operation of the equipment. BSEP LAR Rev 2 Page B-50

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Spurious Actuations Applicability Comments Alignment Alignment Basis Statement Aligns Cables that can cause an undesired spurious actuation are identified by an "S" in the FMEA code section of the circuit information from in FSSPMD. They are evaluated in the SSA in the same manner as "required" cables. The Safe Shutdown Analysis identifies spurious operation candidates which could affect proper safe shutdown system operation. These spurious candidates in this category were addressed by including these devices on the safe shutdown equipment list for the affected safe shutdown system and analyzing them as a safe shutdown component. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.2 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Sectioin 2.1.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 3.0, 9.1.3, 9.3.2 and Attachment 1 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Circuit Information Form Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. BSEP LAR Rev 2 Page B-51

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [B] Associated The concern for the common power source associated circuits is the loss of a safe shutdown power Circuit Cables - source due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on a Common Power non-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordination Source Cables between the upstream supply breaker/fuse feeding the safe shutdown power source and the load breaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss of power to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing its required safe shutdown function. Identify these cables together with the remaining safe shutdown cables required to support control and operation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown. Applicability Comments Alignment Alignment Basis Statement Aligns The guidelines that are used in the evaluation of the common power supplies are as follows:

                             - Using the single-line drawings, ensure that all safe shutdown power supplies required have been included.
                             - For each safe shutdown power supply, review the following documents (as necessary): existing short circuit calculations, load studies, coordination calculations, protective device setting sheets, and time current curves as appropriate to confirm proper coordination between upstream and downstream protective devices to ensure that they are up to date.
                             - In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis shall be considered.
                             - Breaker coordination shall consider both the instantaneous and thermal trip regions of the breaker curves. However, if the analyst must choose between instantaneous and thermal, the instantaneous region of the curve should be the governing consideration.
                             - For cases in which coordination between series protective devices cannot be demonstrated, a common power supply associated circuit are assumed to exist. These circuits are dispositioned by one of the following means:
                             - Demonstrate coordination by refining the available short circuit current and/or trip device characteristics.
                             - Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g., equipment located in same fire area as power supply).
                             - Identify readily achievable protective device setting changes (including changes in fuse size and/or clearing characteristics) that will establish coordination.
                             - Incorporate the Associated Circuits and Cables into the post-fire safe shutdown analysis when protection devices do not provide the desired coordination.
                             - Site specific short circuit and coordination calculations shall be updated as necessary to fully document where coordination is credited for post-fire safe shutdown. The electrical portion of the Safe Shutdown Analysis Report, completed under the Fire Protection Initiatives Project, will include a complete description of the common power supply associated circuits analysis, including reference to applicable supporting calculations and documents.

Comments Reference Document Doc Detail BNP-E-6.085, Unit 2 125/250V DC Coordination/Protection Calculation BSEP LAR Rev 2 Page B-52

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review BNP-E-6.095, Unit 1 125/250V DC Coordination/Protection Calculation BNP-E-8.010, AC Coordination Study BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.1 BNP-E-9.01O0, Safe Shutdown Analysis In Case of Fire Section 3.2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properiy coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [C] Associated The concern with common enclosure associated circuits is fire damage to a cable whose failure Circuit Cables - could propagate to other safe shutdown cables in the same enclosure either because the circuit is Common Enclosure not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result Cables in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple fires). Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown. Applicability Comments Aliqnment Alignment Basis Statement Aligns The following guidelines shall be used in the evaluation of common enclosure associated circuits:

                             - Perform an evaluation of the common enclosure associated circuits by reviewing design and BSEP LAR Rev 2                                                                                                                        Page B-53

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review installation criteria for cable and electrical penetrations. Confirm that cables are adequately protected against short circuits and will not propagate a fire from one fire area to another. In evaluating common power supply circuits the acceptance criteria shall not be limited to standard cable damage temperatures, which are based on not degrading cable insulation (typically 2500C for thermoset cable). Rather, the criteria will be based on not exceeding temperatures at which self ignition or damage to surrounding cables could occur.

                             - If a common enclosure associated circuit is determined to exist, the concern shall be resolved by one of the following means:
                             - Demonstrate by analysis that the cable does not pose a risk to cables within the common enclosure under fault conditions (i.e., the cable exceeds its recommended temperature rise but does not represent a hazard to surrounding cables),
                             - Demonstrate that the lack of fault protection does not adversely affect safe shutdown,
                             - Identify readily achievable protective device setting changes (including changes in fuse size and/or time-current characteristics) that will establish cable protection without affecting other performance requirements, or
                             - Incorporate the cables of concern into the safe shutdown analysis as post-fire safe shutdown cables for the affected power supply.

Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.3 BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire Section 3.2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected BSEP LAR Rev 2 Page B-54

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3 Methodology for Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables Cable Selection and necessary for performing a post-fire safe shutdown analysis. Location Use the following methodology to define the cables required for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis: [Refer to hard copy of NEI 00-01 for Figure 3-41 Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requuirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.1 Identify For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical Circuits Required for diagrams including the following documentation to identify the circuits (power, control, the Operation of the instrumentation) required for operation or whose failure may impact the operation of each piece of Safe Shutdown equipment: Equipment BSEP LAR Rev 2 Page B-55

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review

                             - Single-line electrical diagrams
                             - Elementary wiring diagrams
                             - Electrical connection diagrams
                             - Instrument loop diagrams.

For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concern for the bus under evaluation. If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite or the emergency power source. Aonlicabilitv Comments Alignment Alignment Basis Statement Aligns The safe shutdown component list developed during the safe shutdown analysis sets the stage for the identification of the electrical circuits essential to proper equipment performance. All electrically dependent devices were evaluated in order to identify the corresponding safe shutdown electrical circuits and cables. The circuits identified included power (4.16kV AC, 480V AC, and 125/250V DC), control (120V AC and 125V DC), and instrumentation from the normal operating station (Control Room) and local operating stations. The safe shutdown circuit analysis used one-line diagrams, elementary circuit drawings, and cable block diagrams. Based on the results of this analysis, all of the necessary electrical cables were selected for the subsequent cable routing and separation analysis phases. For each electrical circuit, all circuit cables required for the component to perform as required were identified as being safe shutdown cables and entered into FSSPMD. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. BSEP LAR Rev 2 Page B-56

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.2 Identify In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, Interlocked Circuits cables and equipment. Assign to the equipment any cables for interlocked circuits that can affect the and Cables Whose equipment. Spurious Operation or While investigating the interlocked circuits, additional equipment or power sources may be Mal-operation Could discovered. Include these interlocked equipment or power sources in the safe shutdown equipment Affect Shutdown list (refer to Figure 3-3) if they can impact the operation of the equipment under consideration. Applicability Comments Alignment Alignment Basis Statement Aligns with intent As an alternative to adding the interlocked equipment to the SSEL, it is acceptable to include the cables that are required for the interlocking function (or that could cause the spurious actuation) with the main component that was originally under consideration. Adding them to the components may ease the development of a suitable mitigating strategy in areas where the interlocked cables may be damaged by the fire. Interlocked circuits were either included in the analysis, or the interlocked contact or relay was assumed to be in its worst-case position. Associated circuits identified for each component are either included in the main circuit analysis, or are included by listing the applicable circuit in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" on the Circuit Information Form. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Circuit Information Form Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-57

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance cdteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.3 Assign Cables Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that to the Safe Shutdown may result in maloperation of each piece of safe shutdown equipment. Equipment Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, the same cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component. If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged. Apglicability Comments Alignment Alignment Basis Statement Aligns The circuit analysis for each electrically operated component has been documented on the Circuit Information form within FSSPMD using the input criteria, assumptions, notes, definitions and standard abbreviations described in this section of the procedure. The identification of required cables is not simply a list of cables. It also establishes for each cable a link to the associated component and to the cable's routing and location within the plant. These relationships provide the basis for identifying potential equipment functional failures at a raceway, fire zone, and fire area level. For each electrical circuit, all circuit cables required for the component to perform as required were identified as being safe shutdown cables and entered into the data base management system. Comments BSEP LAR Rev 2 Page B-58

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 and Attachment 1 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Circuit Information Form Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location. NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.4 Identify Identify the routing for each cable including all raceway and cable endpoints. Typically, this Routing of Cables information is obtained from joining the list of safe shutdown cables with an existing cable and raceway database Applicability Comments Alignment Statement Alignment Basis Aligns Cable to raceway information is contained in the Cable Information Forms of the FSSPMD. Cable relationships to SSD equipment and basic events/gates are contained in ARC. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 4.1.2 FSSPMD, Fire Safe Shutdown Program Manager Database Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-002 Closed 08-31-11 El VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. All cable routing used in the NSCA is reference will be applied and the alignment basis will be contained in FSSPMD. This item can be closed. modified as necessary. FRE/Chanae Eval/Mod Corrective Action

Reference:

BSEP LAR Rev 2 Page B-59

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location. NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified. NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.5 Identify Identify the fire area location of each raceway and cable endpoint identified in the previous step and Location of Raceway join this information with the cable routing data. In addition, identify the location of field-routed cable and Cables by Fire by fire area. This produces a database containing all of the cables requiring fire area analysis, their Area locations by fire area, and their raceway. Applicability Comments Alignment Statement Alignment Basis Aligns Cable to raceway information is contained in the Cable Information Forms of the FSSPMD. This includes fire area and zone information. Cable relationships to SSD equipment and basic events/gates are contained in ARC. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 4.1.2 FSSPMD, Fire Safe Shutdown Program Manager Database Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-010 Closed 08-31-11 El VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis / and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. All cable routing used in the NSCA is reference will be applied and the alignment basis will be contained in FSSPMD. This item can be closed. modified as necessary. FRE/Chanqe Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. BSEP LAR Rev 2 Page B-60

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4 Fire Area By determining the location of each component and cable by fire area and using the cable to Assessment and equipment relationships described above, the affected safe shutdown equipment in each fire area Compliance can be determined. Using the list of affected equipment in each fire area, the impacts to safe Assessment shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined. The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document. Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individually mitigating the effects of each of the potential impacts. Applicability Comments Aliqnment Statement Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1 Criteria I The following criteria and assumptions apply when performing fire area compliance assessment to Assumptions mitigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area. Applicability Comments Alignment Statement Alignment Basis N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments BSEP LAR Rev 2 Page B-61

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance cdteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.1 [Number of Assume only one fire in any single fire area at a time. Postulated Fires] Applicability Comments Alignment Alignment Basis Statement Aligns Only a single fire is assumed to occur. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 1.4.1 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.2 and 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.2 [Damage to Assume that the fire may affect all unprotected cables and equipment within the fire area. This Unprotected assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the Equipment and exposure fire that is required by the regulation. Cables] Applicability Comments Alignment Alignment Basis Statement Aligns The following damage was assumed to occur due to a postulated fire: BSEP LAR Rev 2 Page B-62

CP&L Attachment B- NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review

a. Fire damage occurs throughout the fire area under consideration.

b. Fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.3 [Assess Address all cable and equipment impacts affecting the required safe shutdown path in the fire area. Impacts to Required All potential impacts within the fire area must be addressed. The focus of this section is to determine Components] and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected. ADolicabilily Comments Alignment Statement Aliqnment Basis Aligns All potential impacts of the fire are identified in the fault tree. Potential damage to equipment required to show success in each area is addressed with an appropriate compliance strategy. The results are documented in FSSPMD, ARC and in BNP-E-9.006. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.006, Appendix R Separation Analysis Section 3.0 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Attachments 1 and 2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supportinq Detail B2-003 Closed 08-30-11 [] BSEP LAR Rev 2 Page B-63

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Chanqe Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.4 [Manual Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in Actions] accordance with NRC requirements. Applicability Comments Alignment Statement Alignment Basis Statement Aligns Manual actions in support of post-fire shutdown are used to supplement other program elements to ensure post-fire safe shutdown capability. BNP-E-9.007 documents the feasibility of the manual actions. The current regulatory guidance, as reflected in FAQs 06-0012 and 07-0030 was used as the basis for determining the acceptability of the manual actions. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9.006, Appendix R Separation Analysis Section 3.0 BNP-E-9.007, ASSD Manual Action Feasibility BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.2.3.2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.2 and Attachment 2 (para 2.1) (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LARITR Supporting Detail B2-001 Closed 08-23-11 El BSEP LAR Rev 2 Page B-64

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review VFDR Disposition BNP-E-9.007 is the Manual Action Feasibility Study. It BNP-E-9.007 has been completed. has not been completed. Once it is completed add it as a reference. FRE/Change Eval/Mod Corrective Action

Reference:

Date Date Include in VFDR ID Status Entered Due Responsibility LARJTR Supporting Detail B2-004 Closed 08-30-11 11 VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.01 0, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Change EvallMod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.5 [Repairs] Where appropriate to achieve and maintain cold shutdown within 72 hours, use repairs to equipment required in support of post fire shutdown. Applicability Comments Aliqnment Alignment Basis Statement Aligns No repairs are necessary for cold shutdown or to establish safe and stable conditions. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 1.4.6.5 and Attachment 2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Attachment 2 (para 2.1) BSEP LAR Rev 2 Page B-65

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.6 [Assess Appendix R compliance requires that one train of systems necessary to achieve and maintain hot Compliance with shutdown conditions from either the control room or emergency control station(s) is free of fire Deterministic Criteria] damage (III.G.l.a). When cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s):

                             - Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (III.G.2.a)
                             - Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.b).
                             - Enclosure of cable and equipment and associated non-safety circuits of one redundant train within a fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (llI.G.2.c).

For fire areas inside noninerted containments, the following additional options are also available:

                             - Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (IlI.G.2.d);
                             - Installation of fire detectors and an automatic fire suppression system in the fire area (llI.G.2.e); or
                             - Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (IllI.G.2.f).

Use exemptions, deviations and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency depending upon the plant's license requirements. Applicability Comments Alignment Statement Alignment Basis Statement Aligns The NSCA documents the cable/component separation utilized to meet the requirements of NFPA 805. Comments Reference Document Doc Detail BNP-E-9.006, Appendix R Separation Analysis Section 1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Attachments I and 2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.0 BSEP LAR Rev 2 Page B-66

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LARJTR Supporting Detail B2-005 Closed 08-30-11 El VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis! and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Chanae Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.7 [Consider Consider selecting other equipment that can perform the same safe shutdown function as the Additional Equipment] impacted equipment. In addressing this situation, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPP's current licensing basis. Applicability Comments Alignment Statement Aliqnment Basis Aligns This consideration is not clearly stated but is inherent in performing a safe shutdown analysis. BNP-E-9.010 only documents the systems and components that were selected for safe shutdown. The procedure does not specify systems that were considered but not necessary. Comments Reference Document Doc Detail BNP-E-9.01 0, Safe Shutdown Analysis In Case of Fire FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4.2 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-67

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.8 [Consider Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent Instrument Tubing effects on instrument readings or signals associated with the protected safe shutdown path in Effects] evaluating post-fire safe shutdown capability. This can be done systematically or via procedures such as Emergency Operating Procedures. Applicability Comments Aligqnment Alignment Basis Statement Aligns For all credited instruments with instrument sensing lines or equipment that may be supplied by an instrument air line:

1. Identify the instrument sensing lines and/or instrument air lines that are credited in the safe shutdown analysis.

2. Identify the fire zone routing of the individual lines.

3. The sensing and/or air line routing information was entered into the FSSPMD database. These lines are treated in the same manner as cables, and associated with the safe shutdown component.

Equipment ID numbers were developed for the sensing lines that are compatible with PassPort.

4. The sensing lines were incorporated into the fault tree by modeling instrument operation as dependent on sensing line location (If fire occurs in an area where the sensing line is routed, the instrument will be assumed to fail).

5. The instrument air lines were evaluated to determine if they will fail due to a fire in fire areas where instrument air is relied upon to operate. The instrument air lines were incorporated into the fault tree model as necessary.

Sensing lines are treated similar to cables and identified with their routing in FSSPMD. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.7 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). BSEP LAR Rev 2 Page B-68

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.4.2 Methodology for Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area Fire Area assessment. Assessment Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance: [Refer to hard copy of NEI 00-01 for Figure 3-5] Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.1 Identify the Identify the safe shutdown cables, equipment and systems located in each fire area that may be Affected Equipment potentially damaged by the fire. Provide this information in a report format. The report may be by Fire Area sorted by fire area and by system in order to understand the impact to each safe shutdown path within each fire area (see Attachment 5 for an example of an Affected Equipment Report). A~plicability Comments Alignment Alignment Basis Statement Aligns Having identified the location of each component and cable by fire area and using the cable to equipment relationships, the affected safe shutdown equipment in each fire area is determined. Using the list of affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions is determined. Reports are available in ARC. The fault tree is used to determine cascading effects. Comments Reference Document Doc Detail BSEP LAR Rev 2 Page B-69

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.006, Appendix R Separation Analysis Section 3.0 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Attachment 2 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-006 Closed 08-30-11 [] VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Change Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.2 Determine the Based on a review of the systems, equipment and cables within each fire area, determine which Shutdown Paths shutdown paths are either unaffected or least impacted by a postulated fire within the fire area. Least Impacted By a Typically, the safe shutdown path with the least number of cables and equipment in the fire area Fire in Each Fire Area would be selected as the required safe shutdown path. Consider the circuit failure criteria and the possible mitigating strategies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of this assessment since their availability will be important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require this power to support their operation. Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function. Include these cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path. BSEP LAR Rev 2 Page B-70

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipment's position for the specific safe shutdown scenario for the full duration of the shutdown scenario. It is possible for a piece of equipment to be in two different states depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document information related to the normal and shutdown positions of equipment on the safe shutdown equipment list. Applicability Comments Alignment Alignment Basis Statement Aligns By determining the location of each component and cable by fire area and using the cable to equipment relationships, the affected safe shutdown equipment in each fire area can be determined. Using the list of affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area was determined. After the completion of the circuit analyses, cable routing review and development of the CAFTA fault tree model, the ability to bring the reactor to cold shutdown in the event of a fire in any given fire area was evaluated. The evaluation was performed using the CAFTA computer code. CAFTA is instructed to fail all of the safe shutdown components in the fire area where the fire is being postulated. The program then automatically determines the components and systems that fail and if cold shutdown can be achieved and maintained using the fault tree model. If cold shutdown cannot be achieved, safe shutdown strategies were developed. Comments Reference Document Doc Detail BNP-E-9.006, Appendix R Separation Analysis Section 3.0 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 5.1 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LARIT'R Supporting Detail B2-007 Closed 08-30-11 El VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.010, the revised safe shutdown analysis! and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Change Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-71

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.3 Determine Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine Safe Shutdown the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire Equipment Impacts area, and what those possible impacts are. Applicability Comments Alignment Alignment Basis Statement Aligns By determining the location of each component and cable by fire area and using the cable to equipment relationships, the affected safe shutdown equipment in each fire area can be determined. Using the list of affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area were determined. The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in FIR-NGGC-0101. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 9.3 and 9.4 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.4 Develop a The available deterministic methods for mitigating the effects of circuit failures are summarized as Compliance Strategy follows (see Figure 1-2): or Disposition to - Provide a qualified 3-fire rated barrier. Mitigate the Effects - Provide a 1-hour fire rated barrier with automatic suppression and detection. Due to Fire Damage - Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate to Each Required that there are no intervening combustibles within the 20 foot separation distance. Component or Cable - Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability.

                             - Provide a procedural action in accordance with regulatory requirements.

BSEP LAR Rev 2 Page B-72

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review

                              - Perform a cold shutdown repair in accordance with regulatory requirements.
                              - Identify other equipment not affected by the fire capable of performing the same safe shutdown function.
                              - Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process.

Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R section III.G.2.d, e and f. Applicability Comments Alignment Alignment Basis Statement Aligns Compliance strategies for the resolution of component failures that otherwise would have resulted in loss of the ability to ensure that the nuclear safety performance criteria is met for a given fire scenario are maintained in the ARCTM workstation software. Compliance strategies include:

                              "Manual Action Required "Spatial Separation Credited

• Engineering Evaluation
• Repairs Required "Exemption/Deviation "Fire Warp "Radiant Energy Shield "Redundant Trains Available Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.006, Appendix R Separation Analysis BNP-E-9.010, Safe Shutdown Analysis In Case of Fire FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4.2 (NSCA)

Date Date Include in VFDR ID Status Entered Due Responsibility LARrTR Supporting Detail B2-008 Closed 08-30-11 El VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.01 0, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be modified as necessary. FRE/Change Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-73

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.5 Document the Assign compliance strategy statements or codes to components or cables to identify the justification Compliance Strategy or mitigating actions proposed for achieving safe shutdown. The justification should address the or Disposition cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide Determined to each piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-Mitigate the Effects operation could affect safe shutdown, and/or cable for the required safe shutdown path with a Due to Fire Damage specific compliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area to Each Required Assessment Report documenting each cable disposition. Component or Cable APPlicability Comments Alignment Statement Alignment Basis Statement Aligns Compliance strategies for the resolution of component failures that otherwise would have resulted in loss of the ability to ensure that the nuclear safety performance criteria is met for a given fire scenario are maintained in the ARCTM workstation software. Compliance strategies or resolution codes have been assigned to components and/or cables and documented in Separation Analysis Data reports. Comments Reference Document Doc Detail ARC, Fire Safe Shutdown Analysis Workstation BNP-E-9.006, Appendix R Separation Analysis Attachment 1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Attachment 2 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.2 (NSCA) FSSPMD, Fire Safe Shutdown Program Manager Database Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-009 Closed 08-30-11 1l VFDR Disposition The Safe Shutdown Analysis Report is being revised BNP-E-9.01 0, the revised safe shutdown analysis/ and revalidated to include updated Separation Analysis nuclear safety capability assessment (NSCA) was data and information. Once completed the appropriate issued 3/16/12. This item can be closed. reference will be applied and the alignment basis will be BSEP LAR Rev 2 Page B-74

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review modified as necessary. FRE/Chanae Eval/Mod Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5 Circuit Analysis This section on circuit analysis provides information on the potential impact of fire on circuits used to and Evaluation monitor, control and power safe shutdown equipment. Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation. Appendix R Section III.G.2 identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safe shutdown equipment. Section III.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits. Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments BSEP LAR Rev 2 Page B-75

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1 Criteria Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations. Assumptions Applicability Comments Alignment Alignment Basis Statement NIA This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. BSEP LAR Rev 2 Page B-76

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.1 [Circuit Failure Consider the following circuit failure types on each conductor of each unprotected safe shutdown Types and Impact] cable to determine the potential impact of a fire on the safe shutdown equipment associated with that conductor.

                            - A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor. A hot short may cause a spurious operation of safe shutdown equipment.
                            - An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit may prevent the ability to control or power the affected equipment. An open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs]

loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will result in the closure of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial mode of cable failures, are considered to be of very low likelihood. The risk-informed inspection process will focus on failures with relatively high probabilities.

                            - A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductor being applied to ground potential. A short-to-ground may have all of the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to the control circuit or power train of which it is a part.

Consider the three types of circuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area. Applicability Comments Alignment Statement Aliqnment Basis Aligns The circuit analysis shall be reviewed and updated as necessary for credible circuit failures as a deterministic analysis utilizing the Current Design Method (CDM). These failures include:

a. Multiple shorts to ground or grounded conductor.

b. Multiple open circuits.

c. One hot short per affected component or multiple hot shorts for high/low pressure interface components.

Comments BSEP LAR Rev 2 Page B-77

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.B.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.1 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.3 and 9.3.4 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supportina Detail B2-GAP-004 Closed 04-27-12 [E VFDR Disposition EC 85096R0 is evaluating all applicable MOVs for 92- The EVAL-EC has concluded that a specific population 18 issues, inlcuding those thay may be pressure of valves constitute pressure boundary concerns boundary concerns. warranting further engineering analysis, and the follow-on supplemental engineering study is in progress, but it also found that additional data is required to complete the screening. The appropriate follow-up AR tasks have been created to track the continuing evaluations. It is expected that any valve found to be a pressure boundary concern will be considered for modification or replacement, as appropriate. FREIChanae Eval/Mod Corrective Action

Reference:

EC 85096

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected BSEP LAR Rev 2 Page B-78

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.2 [Circuit Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal Contacts and mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst Operational Modes] must consider the position of the safe shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment. Applicability Comments Alignment Alignment Basis Statement Aligns Components are assumed to initially be in their "Normal" operating position (or state) immediately prior to the postulated fire event as identified on the "Circuit Information" form. In most cases the "Normal" position will be the assumed position of the component at 100% power. However, in some cases such as for components that may need to be repositioned due to spurious operation. "Initial" position may differ from the "Normal" position. The component position recorded in the "Normal" position field of FSSPMD is the assumed "Initial" position or state for any circuit analysis on that component. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 4.C FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss BSEP LAR Rev 2 Page B-79

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.3 [Duration of Assume that circuit failure types resulting in spurious operations exist until action has been taken to Circuit Failures] isolate the given circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indicates that fire-induced hot shorts typically self-mitigate after a limited period of time. Aoulicability Comments Alignment SAtegment Alignment Basis Statement Aligns Duration of hot shorts: Industry and NRC cable fire test data indicates that the duration of a hot short is limited. General methodology is to conservatively assume the hot short is maintained (on both AC and DC circuits) until an action is taken to mitigate its affects. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. BSEP LAR Rev 2 Page B-80

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.4 [Cable Failure When both trains are in the same fire area outside of primary containment, all cables that do not Configurations] meet the separation requirements of Section Ill.G.2 are assumed to fail in their worst case configuration. ADplicability Comments Alignment Alignment Basis Statement Aligns The following damage is assumed to occur due to a postulated fire using the deterministic methods:

a. Fire damage occurs throughout the fire area under consideration.

b. Fire damage results in an unusable cable that cannot be considered functional with regard to ensuring proper circuit operation.

Electrical equipment located in a fire area is assumed to fail as a result of the postulated fire in the fire area, and is considered unavailable to ensure completion of safe shutdown functions unless it meets the separation criteria of 10 CFR 50 Appendix R, guidance of NUREG 0800, or is shown to be acceptable as-is based on an approved exemption / deviation. This electrical equipment includes motors, instruments, UiP converters, controllers, switches, MCC's, switchgear, transformers, generators, batteries, panel boards, etc. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss BSEP LAR Rev 2 Page B-81

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5 [A, Circuit The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to Failure Risk identify any potential combinations of spurious operations with higher risk significance. Bin 1 failures Assessment should also be the focus of the analysis; however, NRC has indicated that other types of failures Guidance] required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be followed. Applicability Comments Alignmet Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5 [B, Cable For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the Failure Modes] same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It BSEP LAR Rev 2 Page B-82

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review is reasonable to assume that given damage, more than one conductor-to-conductor short will occur in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between separate cables, commonly referred to as "inter-cable shorting." Inter-cable shorting is less likely than intra-cable shorting. Consistent with the current knowledge of fire-induced cable failures, the following configurations should be considered: A. For any individual multiconductor cable (thermoset or thermoplastic), any and all potential spudous actuations that may result from intra-cable shorting, including any possible combination of conductors within the cable, may be postulated to occur concurrently regardless of number. However, as a practical matter, the number of combinations of potential hot shorts increases rapidly with the number of conductors within a given cable. For example, a multiconductor cable with three conductors (3C) has 3 possible combinations of two (including desired combinations), while a five conductor cable (5C) has 10 possible combinations of two (including desired combinations), and a seven conductor cable (7C) has 21 possible combinations of two (including desired combinations). To facilitate an inspection that considers most of the risk presented by postulated hot shorts within a multiconductor cable, inspectors should consider only a few (three or four) of the most critical postulated combinations. B. For any thermoplastic cable, any and all potential spurious actuations that may result from intra-cable and inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of number. (The consideration of thermoset cable inter-cable shorts is deferred pending additional research.) C. For cases involving the potential damage of more than one multiconductor cable, a maximum of two cables should be assumed to be damaged concurrently. The spurious actuations should be evaluated as previously described. The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research. D. For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multiconductor cable. E. Instrumentation Circuits. Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits. There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly. Applicability Comments Alignment Alignment Basis Statement Aligns Intra-cable Short:

i. For any multiconductor cable (including thermoset, thermoplastic, and armored), any and all potential spurious actuations that may result, including possible combinations of conductors within the cable, may be postulated to occur concurrently regardless of the number.

ii. Intra-cable hot shorts are considered credible events even if the cable is routed within a steel conduit. Inter-cable Short

i. Inter-cable shorting of thermoset cables, or thermoset and thermoplastic cables, are considered to BSEP LAR Rev 2 Page B-83

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review be credible events. ii. For thermoplastic cable, any and all potential spurious actuations that may result from inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of the number. iii. All ungrounded circuit FMEA's are performed postulating that one ground of the opposite polarity already exists due to the fire. Compatible poladty hot shorts for DC circuits were considered to the degree specified in the cases below:

                             - Case 1 - Intra-Cable Shorts within a Single Cable For this case, a single cable must contain both a source and target conductor for both polarities. It is postulated that intra-cable shorts within the cable will result in compatible polarity connections for both polarities (e.g., a plus-to-plus and a minus-to-minus connection for a DC control circuit). Given the relatively high probability of intra-cable conductor-to-conductor shorting, this failure mode was considered.
                             - Case 2 - Intra-Cable Shorts on Separate Cables For this case, two independent but coincident hot shorts of the proper polarity (without grounding) in separate cables must occur. Given the relative high probability of intra-cable conductor-to-conductor shorting, this failure mode was considered.
                             - Case 3 - Inter-Cable Shorts on Separate Cables For this case two independent but coincident hot shorts of the proper polarity (without grounding) must occur. This case differs from Case 1 and 2 in that one or both of the hot shorts must involve inter-cable shorting. Given the low likelihood of coincident proper polarity shorts combined with the low likelihood of inter-cable hot shorting, this failure mode was only considered for components identified as "high-low pressure interface" or Fire PRA "high consequence equipment."

In the plant's review of multiple spurious actuations, the following were considered.

a. Multiple spurious operations resulting from a fire-induced circuit failure affecting a single conductor.

b. Multiple fire-induced circuit failures affecting multiple conductors within the same multi-conductor cable with the potential to cause a spurious operation of a component were assumed to exist concurrently.

c. Multiple fire-induced circuit failure affecting separate conductors in separate cables with the potential to cause a spurious operation of a component must be assumed to exist concurrently when the effect of the fire-induced circuit is sealed-in or latched. There was no specific limit to the number of cables that were considered to be damaged.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.0, 9.3.2, 9.3.3, 9.3.10, 9.4.3, 9.4.6 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment. NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic). BSEP LAR Rev 2 Page B-84

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5 [C, Likelihood Determination of the potential consequence of the damaged associated circuits is based on the of Undesired examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components Consequences] that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown. When considering the potential consequence of such failures, the [analyst] should also consider the time at which the prevented operation or maloperation occurs. Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation. Consideration of cold-shutdown circuits is deferred pending additional research. Applicability Comments Alignment Alignment Basis Statement Aligns As part of changing a plant's Fire Protection licensing basis to meet the requirement of 10 CFR 50.48 (c) and allow for the use of the performance based guidance of NFPA 805, a systematic and complete review of all spurious and multiple spurious operation (MSO) scenarios was required to be performed. The guidance for this review is provided in NEI 00-01, and the process was carried out using an Expert Panel. The Expert Panel was an integral process in the assessment of potential the spurious operations in the post-fire safe shutdown analysis (SSA) and the development of the Fire PRA. The purpose for this MSO Expert Panel was to identify all potential MSO scenarios that could place the plant in an unrecoverable condition, or result in unrecoverable damage to required equipment, and determine which scenarios were credible and may need to be incorporated into the SSA and Fire PRA models. The results of this review have been documented in the BSEP MSO Expert Panel Report, and captured as a standalone document for ease of retrieval, review, and updating as necessary. The Report documents the various scenarios that might result during a fire, and identify those that are valid and should be included in the SSA and Fire PRA models. Scenarios that were screened from inclusion in the plant models are retained in the Report for future reference. Comments Reference Document Doc Detail BNP-0112, BNP MSO Expert Panel Report FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.10 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and BSEP LAR Rev 2 Page B-85

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-0t Ref NEI 00-01 Guidance 3.5.2 Types of Circuit Appendix R requires that nuclear power plants must be designed to prevent exposure fires from Failures defeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, it must be assured that one redundant train of equipment capable of achieving hot shutdown is free of fire damage for fires in every plant location. To provide this assurance, Appendix R requires that equipment and circuits required for safe shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect to the electrical distribution system, the issue of breaker coordination must also be addressed. This section will discuss specific examples of each of the following types of circuit failures:

                             - Open circuit
                             - Short-to-ground
                             - Hot short.

Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety BSEP LAR Rev 2 Page B-86

CP&L Attachment B- NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.1 Circuit This section provides guidance for addressing the effects of an open circuit for safe shutdown Failures Due to an equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit Open Circuit continuity. An open circuit will typically prevent the ability to control or power the affected equipment. An open circuit can also result in a change of state for normally energized equipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV. NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits: Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment. In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evaluate this to determine if equipment fails safe. Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage. Figure 3.5.2-1 shows an open circuit on a grounded control circuit. [Refer to hard copy of NEI 00-01 for Figure 3.5.2-1] Open circuit No. 1: An open circuit at location No. 1 will prevent operation of the subject equipment. Open circuit No. 2: BSEP LAR Rev 2 Page B-87

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not impact the ability to close/stop the equipment. Applicability Comments Aligqnment Alignment Basis Statement Aligns An open circuit may prevent the ability to control or power the affected component. An open circuit may also result in a change of state for normally energized component. For example: the loss of power to a normally open air operated valve's energized solenoid valve due to an open circuit will result in the closure of the valve. For this reason, open circuits are considered in conducting circuit analyses and should consider the following consequences in the circuit analysis when determining the effects of an open circuit:

i. Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required component.

ii. In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of equipment. Evaluate this to determine if equipment fails safe, or if it could cause one or more spurious operations. iii. Open circuit on a high voltage (e.g., 4160V) ammeter current transformer (CT) circuit may result in secondary damage or a fire in the location of the CT itself. The potential CT circuits of concern have been identified, and the final disposition of this potential fire scenario is assessed as part of the SSA/Fire PRA transition to NFPA 805. iv. Shorts-to-ground or short circuits will likely cause a circuit protective device to actuate that results in an "effective" open circuit condition. The analysis shall consider a single open on each conductor in a potentially affected cable of a power circuit. In the case of a control circuit, the analysis considers the combined effects of open circuits if the conductors are contained in the same multi-conductor cable. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA) Date Date Include in VFDR ID Status Entered Due Responsibility LAR/TR Supporting Detail B2-GAP-003 Closed 06-23-12 El VFDR Disposition A more detailed evaluation of the potential for open The evaluation concludes that this failure mode is circuited CTs to cause a secondary fire is required unlikely for CTs that could pose a threat to safe (Section 3.5.2.1). shutdown equipment. FRE/Change Eval/Mod EC 88480 Corrective Action

Reference:

Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-88

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review i*rrp% ovu ow%.;.ul 11MAV41 Q41VLY  %ýC1jJC1L#111LY%011tU1LP%11C11Yb1a0 NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit This section provides guidance for addressing the effects of a short-to-ground on circuits for safe Failures Due to a shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system Short-to-Ground [A, resulting in the potential on the conductor being applied to ground potential. A short-to-ground can General] cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist. Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:

                             - A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment.
                             - In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.

Applicability Comments Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required. Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-89

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit This section provides guidance for addressing the effects of a short-to-ground on circuits for safe Failures Due to a shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system Short-to-Ground [B, resulting in the potential on the conductor being applied to ground potential. A short-to-ground can Grounded Circuits] cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist. Short-to-Ground on Grounded Circuits Typically, in the case of a grounded circuit, a short-to-ground on any part of the circuit would present a concern for tripping the circuit isolation device thereby causing a loss of control power. Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit. [Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-2] Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit. This will result an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected. Short-to-ground No. 2: A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location BSEP LAR Rev 2 Page B-90

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started. Applicability Comments Aliqnment Alignment Basis Statement Aligns The analysis considered a single short-to-ground on each conductor in a potentially affected cable of a power circuit. In the case of a control circuit, the analysis will need to consider the combined effects of shorts-to-ground if the conductors are contained in the same multi-conductor cable. Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit Short-to-Ground on Ungrounded Circuits Failures Due to a Short-to-Ground [C, In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the Ungrounded Circuits] circuit may not result in tripping the circuit isolation device. Another short-to-ground on the circuit or another circuit from the same source would need to exist to cause a loss of control power to the circuit. BSEP LAR Rev 2 Page B-91

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Figure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit. [Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-3] Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the same circuit or on any other circuit fed from the same power source. This will result in an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected. Short-to-ground No. 2: A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started. Applicability Comments Aliqnment Alignment Basis Statement Aligns A single ground fault on an ungrounded AC or DC control circuit has no immediate functional affect. Thus, ungrounded systems are more resilient to functional failures. Nonetheless, multiple ground faults are credible and must be considered. For ease of analysis, an existing - but unspecified - ground fault from the same power source were assumed when analyzing ungrounded systems. Furthermore, multiple shorts-to-ground are evaluated for their impact on ungrounded circuits. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. BSEP LAR Rev 2 Page B-92

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit This section provides guidance for analyzing the effects of a hot short on circuits for required safe Failures Due to a Hot shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between Short [A, General] conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner. Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:

                         - A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.
                         - A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.

Applicability Comments Alignment Alignment Basis Statement Aligns The quality of the hot short connection is important in the ability of any spurious actuation to occur. The hot short circuit failure on the appropriate conductor(s) was conservatively assumed to occur with sufficient electrical contact to impose full voltage on the "target" conductor with a connection that has low impedance. One method for evaluating circuits for the impact of a hot short is using the "hot probe" analysis method. In this approach, the analyst assumes the presence of an energized conductor (the "hot probe") capable of energizing the circuit conductor under consideration. The "hot probe" represents a single "source conductor" without reference to its circuit association (i.e., it could be either a intra-cable or inter-cable source). The hot probe is postulated to make contact with each individual conductor in the cable (separately or coincidentally). For ungrounded control circuits, a slight modification to the hot probe method was used to identify susceptibility to proper polarity, concurrent hot shorts. In this case assume two "hot probes," one representing the positive polarity and one representing the negative polarity. Comments BSEP LAR Rev 2 Page B-93

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit A Hot Short on Grounded Circuits Failures Due to a Hot Short [B, Grounded A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground as Circuits] described above would result in de-energizing the circuit. This would further reduce the likelihood for the circuit to change the state of the equipment either from a control switch or due to a hot short. Nevertheless, a hot short still needs to be considered. Figure 3.5.2-4 shows a typical grounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The following discussion describes how to address the impact of these individual cable faults on the operation of the equipment controlled by this circuit. [Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-4] Hot short No. 1: A hot short at this location would energize the close relay and result in the undesired closure of a BSEP LAR Rev 2 Page B-94

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review motor-operated valve. Hot short No. 2: A hot short at this location would energize the open relay and result in the undesired opening of a motor-operated valve. ADplicability Comments Alignment Statement Alignment Basis Statement Aligns Hot shorts on grounded circuits are considered. Cables susceptible to grounds are identified with the associated equipment. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 and Attachment 1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated. 2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. (a) Common Power Supply Circuits. Those circuits whose fireinduced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device. (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit A Hot Short on Ungrounded Circuits Failures Due to a Hot Short [C, Ungrounded In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spurious Circuits] operation. A single hot short can cause a spurious operation if the hot short comes from a circuit BSEP LAR Rev 2 Page B-95

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review from the positive leg of the same ungrounded source as the affected circuit. In reviewing each of these cases, the common denominator is that in every case, the conductor in the circuit between the control switch and the start/stop coil must be involved. Figure 3.5.2-5 depicted below shows a typical ungrounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The discussion provided below describes how to address the impact of these cable faults on the operation of the equipment controlled by this circuit. [Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-5] Hot short No. 1: A hot short at this location from the same control power source would energize the close relay and result in the undesired closure of a motor operated valve. Hot short No. 2: A hot short at this location from the same control power source would energize the open relay and result in the undesired opening of a motor operated valve. Applicability Comments Alignment Alignment Basis Statement Aligns Hot shorts on ungrounded circuits are considered. Cables suspectible to grounds are identified with the associated equipment. For ungrounded control circuits, a slight modification to the hot probe method is used to identify susceptibility to proper polarity, concurrent hot shorts. In these cases two "hot probes," one representing the positive polarity and one representing the negative polarity are analyzed. All ungrounded circuit FMEA's are performed postulating that one ground of the opposite polarity already exists due to the fire. Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 and Attachment 1 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location. NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be BSEP LAR Rev 2 Page B-96

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review Requirement identified. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.4 Circuit The evaluation of associated circuits of a common power source consists of verifying proper Failures Due to coordination between the supply breaker/fuse and the load breakers/fuses for power sources that are Inadequate Circuit required for safe shutdown. The concern is that, for fire damage to a single power cable, lack of Coordination coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment. For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump. To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump, which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due to the associated circuits' common power source breaker coordination issue, load breaker 4 must be fully coordinated with the feeder breaker to the Train A bus. [Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-6] A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be considered to ensure that coordination is demonstrated at the maximum fault level. The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:

                      - Identify the power sources required to supply power to safe shutdown equipment.
                      - For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus.
                      - For each power source, demonstrate proper circuit coordination using acceptable industry methods.
                      - For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown. Prepare a list of the following information for each fire area:
                       - Cables of concern.
                       - Affected common power source and its path.
                       - Raceway in which the cable is enclosed.
                       - Sequence of the raceway in the cable route.
                       - Fire zone/area in which the raceway is located.

For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate BSEP LAR Rev 2 Page B-97

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review methods. Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables routed in an area of the same path as required by the power source. Evaluate adequate separation based upon the crteria in Appendix R, NRC staff guidance, and plant licensing bases. Applicability Comments Alignment Alignment Basis Statement Aligns Power cables for Safe Shutdown equipment have been selected for evaluation for all components that are required to change states. Coordination of electrical breakers and fuses assure that other power cables from loads on the same electrical bus or distribution center will not adversely impact safe shutdown equipment. Non-SSD load cables required for switchgear operation were also included. This was done to analyze situations where the switchgear loses control power resulting in SSD loads being lost and the loss of the switchgear. FIR-NGGC-0101 also assumes that breaker/fuse coordination exists, but states that if the assumption has not been validated that the cables down stream of the uncoordinated breaker/fuse are added to the circuit analysis. Circuit breaker and fuse coordination are verified by calculations. Comments Reference Document Doc Detail BNP-E-6.085, Unit 2 125/250V DC Coordination/Protection Calculation BNP-E-6.095, Unit 1 125/250V DC Coordination/Protection Calculation BNP-E-8.010, AC Coordination Study BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.1 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.2 and 9.3.6 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location. NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified. NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.5 Circuit The common enclosure associated circuit concern deals with the possibility of causing secondary Failures Due to failures due to fire damage to a circuit either whose isolation device fails to isolate the cable fault or Common Enclosure protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates Concerns along the cable into adjoining fire areas. The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of BSEP LAR Rev 2 Page B-98

CP&L Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns. Applicability Comments Alignment Alignment Basis Statement Aligns An evaluation of common power supply and common enclosure associated circuits was performed. This task verified that the correct cables have been evaluated and that the correct criteria have been selected. If changes are required of any common power supply, or enclosure calculation then fully describe the change made as a result of this post-fire safe shutdown review, or changes required to existing circuits. Circuits not meeting the required criteria shall be dispositioned. A detailed review of protective device curves was performed. For all electrical buses and panels required to support any SSA analysis, a review was made of existing plant calculations to ensure that proper coordination of load and supply breakers on required electrical buses and panels has been performed, and is maintained. This review included all credited AC and DC electrical buses and panels from the highest voltages down to and including 120VAC and 125VDC power distribution panels. EGR-NGGC-0106, "AC and DC Overcurrent Protection and Coordination", provides the methodology used for the selection, setting, and proper coordination of overcurrent protective devices at Nuclear Generation Group (NGG) Plants and establishes the requirements for ensuring proper coordination of breakers. Circuit breaker and fuse coordination are verified by calculations. Adequate coordination exists to assure that a common enclosure issue is not credible. Comments Reference Document Doc Detail BNP-E-9.004, Safe Shutdown Analysis Report Section 3.F.3 BNP-E-9.010, Safe Shutdown Analysis In Case of Fire Section 3.0 FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA) Table B-2 Nuclear Safety Capability Assessment Methodology Review BSEP LAR Rev 2 Page B-99

Enclosure 8 Revised NFPA 805 Transition Report, Attachment J, Fire Modeling V&V

CP&L Affachment J - Fire Modeling V&V CP&L Attachment J Fire Modeling V&V

                                                  -

J. FIRE MODELING V&V 20 Pages Attached Page J-1 BSEP LAR BSEP Rev 22 LAR Rev Page J-1

CP&L Attachment J - Fire Modeling V&V INTRODUCTION This attachment documents the verification and validation (V&V) of the fire models as applied to the Brunswick Fire PRA following the guidance documented in NUREG-1824, and NUREG-1934. These documents are relatively recent joint publications by the US NRC and the Electric Power Research Institute intended to provide guidance on how to conduct and document fire modeling studies, as well as develop the necessary V&V material for supporting these studies. The analysis summarized in this attachment is based on the technical material documented in Report OFP-1212, "Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA". The summary covers all the fire models and the fire modeling applications within the Brunswick Fire PRA as documented in the different calculations prepared for those purposes during the development of the Fire PRA. Each of the models used in the different calculations is identified and a V&V discussion is provided. The report also includes a summary table listing the fire models with the corresponding V&V results. Scope The scope of this study includes the V&V of fire models based on the guidance available in NUREG-1824 and NUREG-1934 as applied in the BSEP Fire PRA. The following subsections list and describe the BSEP fire modeling calculations within the scope of the V&V study. Field Models (Fire Dynamics Simulator, FDS) The BSEP Fire PRA includes two FDS applications, documented in the following calculations: BNP-PSA-080 Rev 2, Attachment 45, BNP-0241, "EDG PPT Reactor and CT Transformer FDS Fire Modeling": This change package documents fire modeling for the PPT Reactor and CT transformers located in the diesel generator building basement. These models are performed to investigate the effectiveness of the thermal shielding provided by the water spray shields installed above these transformers. The fire model within the scope of this validation and verification study is FDS, Version 5. Zone Models (CFAST) The computer model CFAST, Version 6.1.1 is used for the main control room abandonment study documented in BNP-PSA-080 Rev. 3, Attachment 16, Enclosure A (1 RCS04042.103.007-01, Revision 2), Evaluation of Main Control Room Abandonment Times at the Brunswick Nuclear Plant, September 2010. The V&V for CFAST in the BSEP main control room abandonment study is included in the cited report and will be summarized in this document for completeness purposes. Engineering Calculations (Hand Calculations) The BSEP Fire PRA is characterized by a number of engineering (i.e. hand calculations) used throughout the analysis for various purposes. The following subsections provide a brief description of these calculations. Hot Gas Layer Calculations The report BNP-MECH-HGL-001, Rev. 1, "Hot Gas Layer Calculation" documents the approach for determining the damage time for cables immersed in a hot gas layer. The hand calculations used for this analysis are the MQH room temperature correlation for rooms assuming an open BSEP LAIR Rev 2 Page J-2

CP&L Attachment J - Fire Modeling V&V door (NUREG-1805, Chapter 2.6.1) and the Beyler room temperature correlation for closed doors room (NUREG-1805, Chapter 2.6.2). The Attachment 7 to calculation BNP-PSA-080, Rev.1, "Multi Compartment Analysis" describes an analysis for screening multi compartment combinations based on fire modeling. In general, hot gas layer temperatures are calculated for selected fire zones. If the hot gas layer temperature is calculated to be lower than the damage thresholds for cables, the multi compartment scenario is screened. This screening process is based on the hot gas layer calculations performed in BNP-MECH-HGL-0001, Rev. 1. The hand calculations used for this analysis are the MQH room temperature correlation for rooms assuming an open door (NUREG-1805, Chapter 2.6.1) and the Beyler room temperature correlation for closed doors room (NUREG-1 805, Chapter 2.6.2). No multi compartment combinations were screened using this method in BNP-PSA-080 Rev. 3 or Rev. 4. Cable Tray Fire Propagation 9 of BNP-PSA-086, Rev. 1, describes the analysis for determining fire growth profiles in the main control room and the units 1 and 2 cable spreading rooms. The fire growth profiles that are investigated consist primarily of the contribution to the heat release rate from the ignition source and cable trays over a period of time. There is no specific fire modeling model that needs to be subjected to the validation process described in NUREG-1824. The results from this calculation are in the form of heat release rate values generated by:

   " The heat release rate profile assigned to the ignition source, which is an input parameter and not a model requiring validation, and
   " The heat release rate from applicable cable trays, which are calculated following the guidance in Appendix R of NUREG/CR-6850.

The zone of influence (ZOI) elements in this calculation that are applicable (i.e., fire plume conditions) for validation have been addressed in Section 5.4 of Report OFP-1212. HNP-M/MECH-1 194, Rev. 0, "Thermal Damage Time of Cables above a Burning Electrical Cabinet". This calculation describes the approach for determining the time to damage or ignition of the closest cable tray or conduit to an ignition source and subjected to fire plume conditions. The calculation produces a "look up" table for damage or ignition times that are used in the quantification process for calculating non suppression probabilities. The fire model within the scope of this validation and verification study is the Heskestad Plume Temperature Correlation documented in Chapter 9 of NUREG-1805. ZOI Calculations The ZOI calculations in the BSEP Fire PRA are based on hand calculations. These calculations are documented in the following reports:

   " HNP-M/MECH-1 129, Rev. 0, "Fire ZOI Calculation": The goal of this calculation is to calculate ZOI values for various fire sizes that are conservative, encompass a broad set of fuel packages, and integrate more effectively with the scoping fire modeling process.

The fire models within the scope of this validation and verification study are the Heskestad Plume Temperature Correlation documented in Chapter 9 of NUREG-1805 and the solid flame radiation model documented in Section 5.3.2 of NUREG-1805.

   " NED-M/MECH-1007, Rev. 0, "Radiant Energy Target Damage Profile": The purpose of this document is to provide a refinement of the radiant energy ZOI calculation used for identification of transients from electrical cabinet fires. The fire model within the scope of BSEP LAR Rev 2                                                                                Page J-3

CP&L Attachment J - Fire Modeling V&V this validation and verification is the solid flame radiation model documented in Section 5.3.2 of NUREG-1805. NED-M/MECH-1006, Rev. 0, "Generic Fire Modeling Treatments": The generic treatments document offers a set of pre-defined ZOI calculations. A number of fire models are subjected to V&V, which are listed later in Table J-1. Page J-4 BSEP LAR BSEP Rev 22 LAR Rev Page J-4

CP&L Aftachment J - Fire Modeling V&V REFERENCES Industry References

1. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 2 Detailed Methodology," EPRI 1008239 Final Report, NUREG/CR-6850 /

EPRI 1023259, Nuclear Regulatory Commission, Rockville, MD, September, 2005.

2. NUREG/CR-6850 Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," EPRI 1019259, Technical Report, NUREG/CR-6850 Supplement 1, Nuclear Regulatory Commission, Rockville, MD, September, 2010.

3. NUREG-1824, Volume 1, "V&V of Selected Fire Models for Nuclear Power Plant Applications Volume 1: Main Report," NUREG 1824/ EPRI 1011999, Salley, M. H. and Kassawara, R. P., NUREG-1824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., May, 2007.

4. NUREG-1824, Volume 3, "Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 3: Fire Dynamics Tools (FDTS)," NUREG 1824 / EPRI 1011999, Salley, M. H. and Kassawara, R. P., NUREG-1824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., May, 2007.

5. NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide," Salley, M. H. and Kassawara, R. P., NUREG-1934/EPRI-1019195, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Research, Washington, D. C., November, 2012.

6. NUREG-1805, "Fire Dynamics Tools (FDTS)," lqbal, N. and Salley, M. H., NUREG-1805, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., October, 2004.

7. NIST SP 1026, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Technical Reference Guide," Jones, W. W., Peacock, R. D., Forney, G. P.,

and Reneke, P. A., National Institute of Standards and Technology, Gaithersburg, MD, April, 2009.

8. NIST SP 1041, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) User's Guide," Peacock, R. D., Jones, W. W., Reneke, P. A., and Forney, G.

P., National Institute of Standards and Technology, Gaithersburg, MD, December, 2008.

9. NIST SP 1086, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Software Development and Model Evaluation Guide," Peacock, R. D.,

McGrattan, K., Klein, B., Jones, W. W., and Reneke, P. A., National Institute of Standards and Technology, Gaithersburg, MD, December, 2008.

10. NRL/MR/6180-04-8746, "Verification and Validation Final Report for Fire and Smoke Spread Modeling and Simulation Support of T-AKE Test and Evaluation," Tatem, P.A.,

Budnick, E.K., Hunt, S.P., Trelles, J., Scheffey, J.L., White, D.A., Bailey, J., Hoover, J., and Williams, F.W., Naval Research Laboratory, Washington, DC, 2004.

11. Hughes Associates, "Generic Fire Modeling Treatments," Project Number 1SPH02902.030, Revision 0, January 15, 2008.

12. Heskestad, G., "Peak Gas Velocities and Flame Heights of Buoyancy-Controlled Turbulent Diffusion Flames," Eighteenth Symposium on Combustion, The Combustion Institute, Pittsburg, PA, pp. 951-960, 1981.

BSEP LAR Rev 2 Page J-5

CP&L Attachment J - Fire Modeling V&V

13. Heskestad. G., "Engineering Relations for Fire Plumes," Fire Safety Journal, 7:25-32, 1984.

14. Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current,"

Report Number 34, Building Research Institute, Tokyo, Japan, 1960.

15. Yuan, L. and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal, 27, 1996.

16. SFPE, "The SFPE Engineering Guide for Assessing Flame Radiation to External Targets from Pool Fires," Society of Fire Protection Engineers, National Fire Protection Association, Quincy, MA, June, 1999.

17. SFPE Handbook of Fire Protection Engineering, Section 3-1, "Heat Release Rates,"

Babrauskas, V., The SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.

18. NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition"

19. ASME/ANS Ra-Sa-2009, Addenda to ASME/ANS Ra-Sa-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers/American Nuclear Society, New York,

20. NIST SP 1018-5, Volume 2. "FDS Technical Reference Guide Volume 2_ Verification Guide", October 2010.

21. NIST SP 1018-5, Volume 3. "FDS Technical Reference Guide Volume 3_ Validation Guide", October 2010.

Plant Specific References

1. BNP-MECH-HGL-0001, Rev. 1, "Hot Gas Layer Calculation"

2. BNP-PSA-080, Attachment 7, Rev. 1, "Multi-Compartment Analysis"

3. BNP-PSA-086, Attachment 19, Rev. 1, "Cable Tray Fire Propagation"

4. HNP-M/MECH-1 129, Rev. 0, "Fire Zone of Influence Calculation"

5. HNP-M/MECH-1 194, Rev. 0, "Thermal Damage Time of Cables above a Burning Electrical Cabinet"

6. NED-M/MECH-1006, Rev. 0, "Generic Fire Modeling Treatments"

7. NED-M/MECH-1007, Rev. 0, "Radiant Energy Target Damage Profile"

8. BNP-PSA-080 Rev 2, Attachment 45, BNP-0241, "EDG PPT Reactor and CT Transformer FDS Fire Modeling.

9. BNP-PSA-080 Rev 3, Attachment 16, Enclosure A (1RCS04042.103.007-01, Revision 2), Evaluation of Main Control Room Abandonment Times at the Brunswick Nuclear Plant, September 2010.

10. EPM Procedure EPM-DP-FP-001 Rev. 3, Detailed Fire Modeling (Attachment 21 to BNP-PSA-086, Revision 1).

Page J-6 BSEP LAR Rev 22 LAR Rev Page J-6

CP&L Attachment J - Fire Modeling V&V VERIFICATION AND VALIDATION This section includes Table J-1 and Table J-2, which present a summary of the fire models with the corresponding V&V results. Specifically, Table J-1 summarizes the verification and validation results for the different fire modeling calculations listed earlier under the scope section. Table J-2 is specifically devoted to discussing the validation for the fire models used in the generic fire modeling treatment document. The technical material supporting the summary presented in these tables is documented in Report OFP-1212, "Verification and Validation of Fire Models Supporting the Brunswick Nuclear Plant (BNP) Fire PRA". Page J-7 BSEP LAR BSEP Rev 22 LAR Rev Page J-7

CP&L Attachment JJ - Fire Modelinq V&V CP&LAttahmet Fir Moelin V& Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation The ZOI calculations for the Fire Froude number present a few "out of range" results. All of the "out of range" cases are due to calculations exceeding the upper limit of the range, suggesting high intensity fires for the selected fire diameter. One reason for exceeding the upper limit is the use of the 98th percentile heat release rates for the corresponding fire diameters. Based on the guidance in Chapter 8 of NUREG/CR-6850, 98th percentile heat The verification of the models Heskestad release rate values are used for used in support of calculation Use of U.S. Nuclear Plume screening and can be considered on HNP-M/MECH-1129, Rev.0, is Regulatory Temperature the high end of the values assigned to provided in NUREG-1805, Commission (NRC) Correlation ignition sources. In addition, setting the which contains pre-Fire Dynamics Tools documented in Froude number calculation to the upper programmed Microsoft Excel HNP-M/MECH-1 129, 5.1 (FDTs) [NUREG- Chapter 9 of range limit of 2.4 for the 98th percentile Spreadsheets. The Rev 0 1805.0] to determine NUREG-1805 heat release rate values would result in spreadsheets from NUREG-the ZOI of a fire Solid flame a larger diameter. With a larger 1805 are used directly in HNP-scenario in support of radiation model diameter, the flame height calculation M/MECH-1129, Rev.0, scenario development documented in would result in shorter flame lengths, (Attachment 1) and therefore for the BSEP Fire PRA Section 5.2 of and plume temperature calculations additional verification is not NUREG-1805. would suggest lower temperatures. The needed. "out of range" results are based on conservative ZOI calculations for the Fire PRA. Parameters are "in range" for the fire plume application. Parameters are "out of range" for the use of the solid flame radiation model. The reason for number of ZOI results are "out of range" is because the ZOI distances are close to the flames and the experiments selected for validation purposes measured radiation at longer Page J-8 BSEP LAR Rev 22 Page J-8

CP&L Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation distances from the flames. This is a limitation on the available data for validation and not necessarily a limitation on the use of the solid flame radiation model for calculating horizontal components of the ZOI for Fire PRA applications. To account for this limitation, it is noted that validation results from Figure 6-8 in Volume 3 of NUREG-1824 suggest significant heat flux over predictions over the intensity levels used for ZOI calculations (i.e. between 6 and 11 kW/m2) that would result in longer, and therefore conservative, horizontal distances. The ZOI calculations for the Fire Froude number present a few "out of range" The fire modeling documented results. All of the "out of range" cases in this calculation is a are due to calculations exceeding the Microsoft Excel Spreadsheet I upper limit of the range, suggesting supplemented with VBA high intensity fires for the selected fire i Macros. The spreadsheet is a diameter. One reason for exceeding custom built fire modeling the upper limit is the use of the 98th Calculation HNP- Heskestad tools that uses the same percentile heat release rates for the M/MECH-1194, Rev. 0, Plume corresponding fire diameters. Based on HNP-M/MECH-19 Rev./ME '-1194, 52hcdatermages ocrat determines the time Temperature Crelat closed-form plume temperature correlation the guidance in Chapter 8 of Rev. 0 cablesupede docuentin (Chapter 9 of NUREG-1805) NUREG/CR-6850, 98th percentile heat to cables suspended documented in 1 that are provided in NUREG- release rate values are used for over a burning Chapter 9 of i 1824. The calculations screening and can be considered on electrical cabinet. NUREG1805 I included in the workbook have the high end of the values assigned to I been verified to produce the ignition sources. In addition, setting the same results as those Froude number calculation to the upper obtained from the NUREG- irange limit of 2.4 for the 98th percentile 1805, which are verified and . heat release rate values would result in validated in NUREG-1824. a larger diameter. With a larger diameter, the flame height calculation would result in shorter flame lengths, and plume temperature calculations BSEP LAIR Rev 2 Page J-9

CP&L Attachment JJ - Fire Modeling V&V CP&LAttahmen -FireModeinqV& Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation would suggest lower temperatures. The out of range" results are based on conservative ZOI calculations for the Fire PRA. Approximately 60% of the flame length ratio calculations are outside of the range. All of these results are above the high end of the valid range, meaning that the length of the flame was always greater than the height of the target above the fire source causing direct flame impingement on the target. For these cases, given the diameter and HRR of the fire, direct flame impingement occurred to the target cable. Thus, a larger fire diameter would result in flame length ratios within the validation range but lower Fire Froude Numbers. Thus, the values of flame length ratio that are not within the validation range are based on conservative calculations for the Fire PRA. The calculation also describes how the time to target damage is determined using the damage time tables available in Appendix H of NUREG/CR-6850 for Thermoset or Thermoplastic cables. The technical basis for the time calculations is available in Appendix H of NUREG/CR-6850 and is not addressed in this V&V study as there are no dimensionless parameters for time to damage to be evaluated. HNP-M/MECH-1194, Rev. 0, provides justification for the conservative use of the damage time tables in Appendix H of NUREG/CR-6850. Page J-1O BSEP LAIR Rev 22 LAR Rev Page J-1 0

CP&L Aftachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation The comparison of dimensionless parameters with the validation range suggest a number of "out of range" results, which are expected for both the Fire Froude Number and the radial distance ratio dimensionless parameter. i To ensure the equations were Fothraildsncrto For the radial distance ratio T codensorec the equations e dimensionless parameter, all the Calculation NED- coreadseed cr set in the calculations that are "out of range" are

                                                                                      -spreadsheets used in the          on the low side of the range. This M/MECH-11007, Rev. 0,                           calculation, the spreadsheet       happens because the target is close to utilizes the Nuclar U.S.        Rgulaorythe            results  wereofchecked results   the NUREG-against   the flames and the experiments Nuclear Regulatory           Solid flame        t                                  selected for validation purposes Commission Dynamics(NRC)          Cradiation model 21805 NE-/EH10,Fire                                Tools                                  FDTs Solid spreadtion    for Flame Model   mesue fnr               m                          ad longendisance distances NED-M/MECH-1007,       5.3                                         documented in                                         of the target from the flames. This is a Rev. 0                                (FDTs) [NUREG-               Section 5.2 of     inputs. Both spreadsheet           limitation on the available data for 1805.0] to determine a       NUREG-1805.        models were found to produce            lidation and not necessarily a radiative ZOI from           N                  the same results (NED-                limation on t ueoesoila electrical cabinet fires                        M/MECH-1007, Rev. 0, page            iation ode for calculaing to qualified and                                11), therefore the                 i       adi omonenof             theuZaifo unqualified cables.                             spreadsheets used in the           horizont calculations are considered        Fire PRA applications. To account for verifiedthis                               limitation, it is noted that validation results, from Figure 6-8 in Volume 3 of NUREG-1 824, suggest significant heat flux over predictions over the intensity levels used for ZOI calculations (i.e.,

between 6 and 11 kW/m2) that would result in longer, and therefore conservative, horizontal distances. Attachment 19 of BNP- The fire modeling documented In practice, the purpose of this PSA-080, Rev 1 Plume in this calculation is conducted calculation is to credit cable tray covers describes the analysis temperature in a modified version of the for preventing fire propagation through for determining fire models EPM fire modeling workbook . cable trays per the guidance in Attachment 19, Rev. growth profiles in the described des cribed in escribed in detail in Appendix Q of NUREG/CR-6850, which main control room and HNP-M/MECH- EPM Procedure EPM-DP-FP- provides the technical basis for applying the units 1 and 2 cable 1194, Rev. 0 1001, Rev. 3. The EPM fire such credit. Consequently, there is no spreading rooms. The modeling workbook is a I specific fire model that needs to be fire growth profiles that custom built fire modeling tool 1 subjected to the validation process firePgrowtRepro2iPagethati BSEP LAIR Rev 2 Page J- 11

CP&L Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation are investigated consist that uses the same closed- described in NUREG-1824. The results primarily of the form correlations that are i from this calculation are in the form of contribution to the heat provided in the V&V'd heat release rate values generated by: release rate from the chapters of NUREG-1824. The heat release rate profile assigned ignition source and The tool has been modified for to the ignition source, which is an input cable trays over a this activity to include only parameter and not a model requiring a period of time. those portions of the workbook 1 Fire Model Validation process, and necessary to calculate a ZOI for each analyzed ignition The heat release rate from applicable source. The calculations I cable trays, which are calculated included in the workbook have following the guidance in Appendix R of been verified to produce the NUREG/CR-6850. same results as those obtained from the NUREG-1805, which are verified and validated in NUREG-1824. Calculation MECalcao BNP- -0, The handThe Beyler room temperature MECH-HGL-0001, Th cauans The fire modeling documented I correlation was developed using data Rev.1, determines the calculations in this calculation is a with a maximum temperature rise of fire heat release rate used for this Microsoft Excel Spreadsheet 1150°C. Extrapolation of this correlation necessary to generate analysis are the supplemented with VBA to higher temperatures (330*C) is layer within a temighga M re temperature Macros. The spreadsheet is a justified by using the Beyler correlation compartment for a correlation for custom built fire modeling tool only when it is the most conservative comaten flort arrreton f that uses the same closed- result (i.e., lower estimate of HRR for given floor area. rooms form room temperature room-wide damage to cables), Furthermore, this s BNP-MEClulatidesorie, assuming an BNP-MECH-HGL- calculation describes o open dr mng door correlations (Sections 5.1 and i compared to the MQH correlation, 0001, Rev. 1 5.5 the process for (NUREG-1805, 5.3 of NUREG-1805) that are which is validated at higher crediting thecredtingthe hea "heat Chater Chapter 2.1).1) in the V&V'd provided of NUREG-1824. chapters 1 temperatures soak" time. The "heat and the Beyler c The results show that more than 90% of soak totie soak" time refers herfersto the oomthe room Theworkbook calculationshave been in included 1 the compartment ratio parameters are lag time between the temperature te within the valid range, suggesting that gtemperetu temperature e cosedadoorfor correlation verified to produce the same results as those obtained from the room size of these fire scenarios was included in the V&V study surrounding the cable closed doors the NUREG-1805, which are wadescribed in NUREG-1824. Those targets and the room (NUREG- verified validated inos in temperatures inside the 1805, Chapter reand validated in compartment aspect ratios that fall tempertatrges insie toutside the application range do so on cgenerating the both ends of the range. This can be geEnerRaRen2gPaehe-1 BSEP LAR Rev 2 Page J-1 2

CP&L Aftachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation electrical damage I explained by the limited experiments and/or ignition. selected for the validation study. As indicated in NUREG-1934, the selected experiments are representative of various types of spaces in commercial NPPs, but do not encompass all possible geometries or applications. This is a limitation on the available data for validation and not necessarily a limitation on the use of the model for calculating HGL scenarios applicable to the Fire PRA. To address this limitation, it is noted that both the MQH and Beyler room temperature models are reported to overpredict room temperatures for most configurations in Table 3-1 of NUREG-1824, Volume 1 (which lists a yellow-plus) and Table 4-1 in NUREG-1934 (which suggests an average bias of 1.44). This over prediction throughout the evaluated scenarios suggest that the configurations that are outside the 1 validation range in this application will also result in temperature over predictions. Calculation BNP-PSA- The hand The fire modeling documented The results show that more than 80% of 080, Attachment 7, calculations in this calculation is a the compartment ratio parameters are Rev. 3, determines the used for this Microsoft Excel Spreadsheet within the valid range, suggesting that fire heat release rate analysis are the supplemented with VBA i the room size of these fire scenarios necessary to generate MQH room Macros. The spreadsheet is a was included in the V&V study BNP-PSA-080, 5.6 a damaging hot gas temperature custom built fire modeling described in NUREG-1824. The Attachment 7, Rev. 3 layer within a multi- correlation for tools that uses the same remaining compartment aspect ratios compartment rooms closed-form room temperature 1 fall outside the high end of the range. combination for a assuming an correlations (Sections 5.1 and This can be also explained by the combined floor area. open door 5.3 of NUREG-1805) that are limited experiments selected for the These results were (NUREG-1805, provided in the V&V'd validation study. As indicated in generated for a Chapter 2.1) chapters of NUREG-1824. NUREG-1934, the selected BSEP LAR Rev 2 Page J-13

CP&L Aftachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation screening criterion that and the Beyler The calculations included in experiments are representative of was not implemented in room the workbook have been various types of spaces in commercial BNP-PSA-080, temperature verified to produce the same NPPs, but do not encompass all Attachment 7, Rev. 3 or correlation for results as those obtained from possible geometries or applications. Rev. 4 closed doors the NUREG-1805, which are This is a limitation on the available data room (NUREG- verified and validated in for validation and not necessarily a 1805, Chapter NUREG-1824. limitation on the use of the model for 2.3). calculating HGL scenarios applicable to the Fire PRA. To address this limitation, it is noted that both the MQH and Beyler room temperature models are reported to overpredict room temperatures for most configurations in Table 3-1 of NUREG-1824, Volume 1 (which lists a yellow-plus) and Table 4-1 in NUREG-1934 (which suggests an average bias of 1.44). This over prediction throughout the evaluated scenarios suggest that the configurations that are outside the validation range in this application will also result in temperature over predictions. The "Generic Fire Modeling Treatments," Revision 0 document is The calculation development used to establish ZOI and review proess in place atl for specific classes of the time the 'Generic Fire NED-M/MECH-1006, ignition sources and Listed in Table Modeling Treatments" Rev. 0 5.7primarily serves as a later in this document was prepared See Table later in this section. screening calculation in section. included contributions from a the Fire PRA under calculation preparer, a NUREG/CR-6850 calculation reviewer, and a Sections 8 and 11. calculation approver. Page J-14 BSEP LAR Rev 2 LAR Rev 2 Page J-14

CP&L Attachment JJ - Fire Modeling V&V CP&LAttahmet Fir Moelin V& Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation The fire model FDS Version 5 has been verified and validated, and the V&V studies are documented and available as NIST reports. However, the specific scenario configuration has The fire modeling analysis was characteristics that are not included in conducted using FDS Version the fire experiments used for validation

5. This version was released purposes. To address this limitation, after NUREG-1824 Volume 7 the conclusions suggested by the FDS The purpose of this was published. However, simulation have been compared with calculation is to NIST developed V&V studies hand calculations for fire plume and evaluate plume for FDS version 5 which are flame radiation. The comparison temperatures documented in the following suggests that the FDS results indicating immediately above reports: no plume damage to targets from a 69 metal water spray kW fire given the obstructions (metal covers available above NIST SP 1018-5, ,FDS water spray covers above the selected transformers transformers) are consistent with the BNP-0241, Technical Reference Guide to identify if cable hand calculation results, which Attachment 45 to 5.8 FDS Version 5 Volume 2_ Verification Guide targets can be assumes unobstructed plumes. The BNP-PSA-080 Rev 2 damaged. The metal hand calculations suggests water spray covers are NIST SP 1018-5, FDS unobstructed plume temperature that installed above the Technical Reference Guide can be damaging close to the ZOI limit, transformers to protect Volume 3_ Validation Guide from what is concluded that the them from sprinkler These V&V guides follow the obstruction should provide protection to spray in the event of an same structure and technical the target by breaking the plume at that inadvertent actuation of approach as NUREG-1924 location. Furthermore, the ignition the sprinkler system. and provide the verification sources are relatively large metal necessary for supporting the enclosures that will provide further use of FDS Version 5 for plume obstructions for fires postulated commercial nuclear power inside the enclosures (i.e. inside the plant applications. transformers).

Page J-15 BSEP LAR Rev 22 BSEP Page J-15

CP&L Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific BSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation Appendix D, Section D.1 of This calculation report 1RCS04042.103.007-01 The non-dimensional parameters that documents the includes a software description affect the model results, as documented dvauatont temand benchmark V&V. The in NUREG-1824, Volumes 1 and 5 and verification for the CFAST NUREG-1934, include the model BNP-PSA-080 Rev 3, control room model (Version 6.1.1) is geometry, the global equivalence ratio, Attachment 16, abandonment times. CFAST, provided in NUREG-1824, the fire Froude Number, and the flame EnRClSure42.1A . tes abandoment uVersion 6.1.1 Volume 5. Supplemental length ratio. Non-dimensional (1 Rev2)timets ae then used averification for CFAST, parameters that relate to target 01 Rev 2) inputs to the risk quantification of main Version 6.1.1 is provided as exposure conditions (heat flux) and control room fire an attachment to sprinkler actuation (ceiling jet) are not scenarios. 1RCS04042.103.007-01 as applicable to this calculation because well as in NIST SP 1086 (Ref. these output parameters are not used. 9). Page J-16 BSEP LAR Rev Rev 22 Page J-16

CP&L Attachment J - Fire Modeling V&V Table J-2: V &V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations. Location in Reference Generic Fire in"Generic Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Treatments" Verification* (Ref. 11) (Ref. 11) Document* Flame Height Page 18 Heskestad Provides a limit Directly (Ref. 19); on the use of the cpT" .*..] NUREG-1824, 5 Heskestad ZO -05 < lgl° 0 ZD5

                                                                                                   <   Volume 3 (Ref.

(Ref. 20) 23) 4nAH, 2 < 3000

                                                                                                                           ,lrD Indirectly In practice, wood and hydrocarbon      NUREG-1824, fuels, momentum or buoyancy            Volume 5 (Ref.

dominated, with diameters between 10) 0.05 - 10 m (0.16 - 33 ft). (Correlation used in CFAST) Point Source Page 19 Modak (Ref. Lateral extent of Isotropic flame radiation. Compared NUREG-11824, Predicted heat flux at Model 45) ZOI - with data for 0.37 m (1.2 ft) diameter Volume 3 (Ref. target is less than 5 comparison to PMMA pool fire and a target located 23); kW/m2 (0.4 4 Btu/s-ft2) per other methods at a L ratio of 10. SFPE (Ref. 24) SFPE. R Method of Page 19 Shokri et al. Lateral extent of Pool aspect ratio less than 2.5. SFPE (Ref. 24) Ground based vertical Shokri and (Ref. 46) ZOI - Hydrocarbon fuel in pools with a NUREG-1824, target. Beyler comparison to diameter between 1 - 30 m (3.3 - 98 Volume 3 (Ref. other methods ft). 23) Vertical target, ground level. Method of Page 20 Mudan (Ref. Lateral extent of Round pools; SFPE (Ref. 24) Total energy emitted by Mudan (and 47) ZOl - Hydrocarbon fuel in pools with a thermal radiation less Croce) comparison to diameter between 0.5 - 80 m (1.64 - than total heat released. other methods 262 ft). Page J-17 BSEP BSEP LAR Rev 2 Page J-17

CP&L Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations. Location in inReference "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Treatmentsg Verification* (Ref. 11) (Ref. 11) Document* Method of Page 20 Shokri et al. Lateral extent of Round pools; SFPE (Ref. 24) Predicted heat flux at Shokri and (Ref. 46) ZOI Hydrocarbon fuel in pools with a target is greater than 5 2 Beyler diameter between 1 - 50 m (3.3 - NUREG-1824, kW/m2 (0.44 Btu/s-ft )per 164 ft). Volume 3 (Ref. SFPE (Ref. 24).

23) Shown to produce most conservative heat flux over range of scenarios considered among all methods considered.

Plume heat Page 22 Wakamatsu Vertical extent of Fires with an aspect ratio of about 1 Wakamatsu et Area source fires with fluxes et al. (Ref. ZOI and having a plan area less than 1 al. (Ref. 48) aspect ratio - 1. Used

48) m2 (0.09 ft 2). (larger fires) with plume centerline SFPE Handbook temperature correlation; of Fire Protection most severe of the two is Engineering, used as basis for the ZOI Section 2-14 dimension. This is not a (Ref. 49) constraint in the fire model analysis for the cases evaluated.

Plume Page 23 Yokoi (Ref. Vertical extent of Alcohol lamp assumed to effectively NUREG-1824, Area source fires with centerline 21); ZOI be a fire with a diameter -0.1 m (0.33 Volume 3 (Ref. aspect ratio - 1. Used temperature Beyler (Ref. ft). 23); with plume flux

50) SFPE Handbook correlation; most severe of Fire Protection of the two is used as Engineering, basis for the ZOI Section 2-1 dimension.

(Ref. 51) Page J-18 BSEI' LAR Rev 2 BSEP Page J-18

CP&L Attachment J - Fire Modeling V&V Table J-2: V &V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations. Location in Reference in "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments (ref. Treatments" Treatments" Verification* (Ref. 11) (Ref. 11) Hydrocarbon Page 51 SFPE Determine heat Hydrocarbon spill fires on concrete None. Based on None. Transition from spill fire size Handbook of release rate for surfaces ranging from -1 to -10 m limited number of unconfined spill fire to Fire unconfined (3.3 - 33 ft) indiameter. observations, deep pool burning Protection hydrocarbon spill assumed to be abrupt. Engineering, fires. Section 2-15 (Ref. 52) Flame Page 100 SFPE Determine the Corner fires ranging from -10 to None. Based on None. Offset is assumed extension Handbook of fire offset for -1,000 kW (9.5 - 948 Btu/s). Fires limited number of equal to the depth of the Fire open panel fires. included gas burners and observations, ceiling jet from the Protection hydrocarbon pans. experiments. Engineering, Section 2-14 (Ref. 53) Line source Page 101 Delichatsios Determine the Theoretical development. SFPE Handbook None. Transition to area flame height (Ref. 54) vertical extent of of Fire Protection source assumed for the ZOI Engineering, aspect plan ratios less Section 2-14 than four. Maximum of (Ref. 49) area and line source predictions used in this region. Corner flame Page 108 SFPE Determine the Corner fires ranging from -10 to None. None. height Handbook of vertical extent of -1,000 kW (9.5 - 948 Btu/s). Fires Correlation form Fire the ZOI included gas burners and is consistent with Protection hydrocarbon pans. other methods; Engineering, comparison to Section 2-14 dataset from (Ref. 53) SFPE Handbook, Section 2-14 (Ref. 53) provides basis. BSEP LAR Rev 2 Page J-19

CP&L Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations. Location in inReference "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Moeaeng Verification* (Ref. 11) (Ref. 11) Document* Air mass flow Page 140 Kawagoe Compare Small scale, %scale, and full scale Drysdale (Ref. None. SFPE (Ref. 57) through (Ref. 55) mechanical single rooms with concrete and steel 56); spaces with a wide range opening ventilation and boundaries. Vent sizes and thus SFPE (Ref. 57) of opening factors. natural opening factor varied. Wood crib ventilation fuels. Line fire flame Page 210 Yuan et al. Provides a limit None. None. height (Ref. 22) on the use of the Z Correlation form ZOI (ZOI); 0.002 < T < 0.6 is consistent with Extent of ZOI for other methods; cable tray fires. In practice, from the base to several comparison to times the flame height based on dataset from 0.015 - 0.05 m (0.05 - 0.16 ft) wide Yuan et al. (Ref. gas burners. 22) provides basis. Cable heat Page 210 NBSIR 85- Provides Cables with heat release rates per None. Correlation predicts a release rate 3196 (Ref. assurance that unit area ranging from about 100 - lower heat release rate per unit area 58) the method used 1,000 kW/m2 (8.8 - 88 Btu/s-ft2). than assumed in the is bounding Treatments and is based on test data. Line fire plume Page 212 Yuan et al. Provides a limit None. None. centerline (Ref. 22) on the use of the Z Correlation form temperature ZOI (ZOI); 0.002 < T < 0.6 is consistent with Extent of ZOI for other methods; cable tray fires. In practice, from the base to several comparison to times the flame height based on dataset from 0.015 - 0.05 m (0.05 - 0.16 ft) wide Yuan et al. (Ref. gas burners. 22) provides basis. Page J-20 BSEP LAR Rev Rev 2 2 Page J-20

CP&L Attachment J - Fire Modeling V&V Table J-2: V &V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations. Location in Reference in "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling odeling Application Original Correlation Range Validation and Modeling Treatments Mnti Treatments Creeatn Verification* (Ref. 11) (Ref. 11) Treatments" Ventilation Page 283 Babrauskas Assessing the Ventilation factors between 0.06 - SFPE (Ref. 57) None. Provides depth in limited fire size (Ref. 59) significance of 7.51. the analysis of the vent position on Fire sizes between 11 - 2,800 kW selected vent positions. the hot gas layer (10 - 2,654 Btu/s) The global equivalence temperature Wood, plastic, and natural gas fuels. ratio provides an alternate measure of the applicability of the analysis and for reported output is within the validation range of CFAST.

• Reference number provided in parentheses in Table J-2 refers to the reference number in Generic Fire Modeling Treatments (Ref. 11)

Page J-21 BSEP BSEP LAR Rev Rev 2 2 Page J-21

Enclosure 9 Revised NFPA 805 Transition Report, Attachment L, NFPA 805 Chapter3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval CP&L Attachment L NFPA 805 Chapter 3 Requirements for Approval

                                     -

L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii) 11 Pages Attached Page L-1 BSEP LAR BSEP Rev 22 LAR Rev Page L-1

CP&L Attachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 1 NFPA 805 Section 3.5.16 NFPA 805 Section 3.5.16 states: "The fire protection water supply system shall be dedicated for fire protection use only. Exception No. 1: Fireprotection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclearsafety flow demands for the duration specified by the applicable analysis. Exception No. 2: Fireprotection waterstorage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determinedin this section." Contrary to the requirements of NFPA Section 3.5.16, BSEP utilizes Fire Protection Water in the following plant support applications: o Containment Heat Removal. In the event that nuclear service water is lost to the RHR heat exchangers, the WFSS may be used to provide backup cooling for containment heat removal. This is detailed in procedure 0AOP-18.0. o Coolant Injection. In the event of a failure of the normal reactor level control systems to maintain water level, the WFSS may be used as an alternate coolant injection system. This is detailed in procedure 0EOP-01-LEP-01. In addition fire water may be used for alternate boron injection. This is detailed in procedure OEOP-01-LEP-03 o Fuel Pool Cooling. Fire hoses on the Reactor Building 117' elevation may be used as a makeup water source if the spent fuel pool level cannot by recovered by normal means. This is detailed in procedure OAOP-38.0. o RHR Service Water Shutdown and wet layup process. This is detailed in procedure 1(2)OP-43. o Flushing/filling. Venting RHR service water and heat exchangers in accordance with 1(2)OP-43. o RHR Service Water System Operability Test. This is detailed in procedure OPT-08.1.4a(b). o Flushing Radwaste Rad Monitor. This is detailed in procedure OOP-06.4. o Seal water to Storm Drain Collector Basin Pumps. This is detailed in procedure OOP-54. o Temporary Cooling Water Supply to Service Air Compressor 1(2)D. This is detailed in procedure OOP-46. o Refill of SBGT drain trough. This will be detailed in procedure 001-03.3 (PRR-553262). Page L-2 BSEP Rev 2 LAR Rev BSEP LAR 2 Page L-2

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval Basis for Request: The basis for use of fire protection water supply for emergency uses is discussed below.

1. Containment Heat Removal The Nuclear and Conventional Service Water Systems are used for removing heat from the RHR system, Diesel Generator, and Reactor Building Closed Cooling Water (RBCCW) systems. The RHR system in turn is used for removing heat from the Primary Containment and for reactor core decay heat removal.

In the unlikely event of a complete and sustained loss of Nuclear and Conventional Service Water, Abnormal Procedures direct the operator to align water from the fire protection tank to the RHR Heat exchangers. This would be done only after attempts to restore service water flow from any one of five pumps are unsuccessful and if actions to isolate major service water system leaks are not successful. Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly. Simultaneous use with the other three emergency uses is not likely. Estimated flow and pressure demand for the Electric and Diesel Fire Pumps is 2000 gpm each, at a normal system operating pressure of approximately 125 psig. (Ref. 0AOP-18.0)

2. Alternate Coolant Injection In the unlikely event that reactor water level cannot be restored and maintained using installed high and low pressure injection systems, the Emergency Operating Procedures direct the operator to restore reactor coolant level using all of the following systems: Standby Liquid Control, Heater Drains, Service Water, Demineralized Water, and Fire Protection Water. Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly. Simultaneous use with the other three emergency uses is not likely.

Estimated flow and pressure demand for the Electric and Diesel Fire Pumps are 2000 gpm each, at a normal system operating pressure of approximately 125 psig. (Ref. 0EOP-01-LEP-01)

3. Alternate Boron Injection Emergency Operating Procedures direct the operator to inject boron if it has been determined that the reactor will not remain shutdown under all conditions without boron and following a reactor scram. The operator is directed to inject boron with one or more of the following systems: Control Rod Drive, HPCI, RCIC, and RWCU. If RWCU System is used, the emergency procedure directs the operator to use a 1-1/2 inch fire hose to fill the system precoat tank to pre-mix boron for injection. Estimated flow from the fire house should be less than 100 gpm, at a normal system operating pressure of approximately 125 psig. Filling a BSEP LAR Rev 2 Page L-3

CP&L Attachment L - NFPA 805 Chapter 3 Requirements for Approval completely empty precoat tank should require not more than 300 gallons. There is adequate margin for this use concurrent with fire suppression. Simultaneous use with the other three emergency uses is not likely. (Ref. OEOP-01-LEP-03)

4. Fuel Pool Cooling The spent fuel pool decay heat removal systems consist of Fuel Pool Cooling and Alternate Decay Heat Removal System. Additionally the Division II RHR system is capable of providing fuel pool cooling assist. If an abnormal event occurs that results in decreasing water level or increasing fuel pool temperature, the following systems will be aligned to provide make up and cooling. In the unlikely event of a complete and sustained loss of these systems, Abnormal Operating Procedures direct the operator to add water from the Demineralized Water System and the Fire Protection Water System. Fire hoses will be used to direct as much water as necessary to restore and maintain water level in the spent fuel pool. Flow for three fire hoses is estimated at 250 gpm, at a normal system operating pressure of approximately 125 psig. Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly.

Simultaneous use with the other three emergency uses is not likely. Following the use of fire protection water for Containment Heat Removal, Coolant Injection, Alternate Boron Injection, or Fuel Pool Cooling, fire protection water restoration requirements would be to realign the system to standby and refill the Fire Protection Tank from the Brunswick County Water Supply System. Restoration is described in plant operating procedures. The Fire Protection Water Tank is filled and maintained through 1.5 -inch and 4-inch air-operated fill valves. Level switches on the tank control automatic makeup, and an annunciator alert is provided in the MCR if the tank level is drawn down. Two alarms are provided at a low and a low-low level alarm set point. Routine surveillance checks by plant operators using a local tank level indicator verify that the tank level is kept above the minimum level. Level instrumentation is in feet above the tank bottom. A manual bypass valve may also be used to refill the tank. Fill water is supplied by a 15,000 gallon on site County Water Storage Tank, with two parallel pumps supplying flow. The County Water Storage Tank is maintained full by the Brunswick County Water Main by two pumps delivering 200 gpm each. Additionally, a design feature is to manually align the Electric Motor and Diesel driven fire pump suctions to the Demineralized Water Storage Tank, which will allow time for the Fire Protection Storage Tank to refill. The Demineralized Water Storage Tank is checked regularly by Operations, in the same manner as the Fire Protection Water Storage Tank, to verify it is 14 foot above the tank bottom. Use of fire protection water for Containment Heat Removal, Coolant Injection, Alternate Boron Injection, and Fuel Pool Cooling are strictly controlled by Emergency Operating and Abnormal Operating procedures. Operators are trained regularly on these procedures and the equipment. Communications are by the plant Public Address (PA) System, sound powered phones, and operations radio systems. BSEP LAR Rev 2 Page L-4

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval The basis for use of fire protection water supply for non-emergency uses is discussed below. Based on flow rates and volumes used from the fire protection water supply explained in normal evolutions one through seven below, the margin available in the fire protection water supply system is adequate. The largest water demand for a safety related area, Unit 2 South RHR, and the largest water demand for a non-safety related area, the Main Transformer, were both calculated to be within the capacity of a single fire pump. These flow demands are discussed in DBD-62, Section 3.3.4. Estimated flow, pressure, and expected frequency, as applicable, are discussed in the paragraphs below. Each normal evolution is performed under procedural controls. Annunciators for low tank level or local monitoring will alert the operator such that minimum tank level will not be violated. Alerts are provided by tank low and low-low alarms, along with Electric Motor and Diesel Driven Fire Pump running annunciators in the MCR.

1. Residual Heat Removal (RHR) Service Water Shutdown and Wet Layup Process Usage of fire protection water for RHR Service Water (RHRSW) wet layup should be allowed because there is no appreciable flow of fire water from the Fire Water Storage tank. Wet layup following RHRSW system shutdown does not place a significant drain on the Fire Protection system. The RHRSW automatic valve controls and operating procedures will isolate valve 1(2)SW-V143 if the RHR Service Water system is placed in service to the RHR heat exchangers. While in a static wet layup alignment any RHR Service Water system leakage should not exceed the capacity of the county water make up to the Fire Protection Water storage tank or the capacity of the two fire pumps.

2. Flushing, filling, and venting RHR Service Water and Heat Exchangers Usage should be allowed because procedural controls prevent the operator from lowering tank level below the low level alarm set point. When performing the flush, operating procedures require an operator to be stationed to continuously monitor tank level locally and to maintain direct communications with the MCR by plant PA or radio. Procedures require usage of not more than one-half foot tank level for each flush evolution. This amount of usage is well within the capacity of the county water makeup flow.

3. RHR Service Water System Operability Test Quarterly testing is performed on each RHRSW system. There are two divisions on each unit. Following each test the system is flushed per operating procedures. Not more than one-half foot of Fire Protection Storage Tank level is used for each flush evolution. Usage is well within the Fire Protection Water Storage Tank makeup capacity and the volume stored in the County Water Storage Tank is more than enough for immediate use. Usage should be allowed BSEP LAR Rev 2 Page L-5

CP&L Attachment L - NFPA 805 Chapter 3 Requirements for Approval because procedural controls prevent the operator from lowering tank level below the low alarm set point and immediate makeup is available.

4. Flushing Radwaste Radiation Monitor Periodic flushing of the Radwaste Radiation Monitor is performed at an estimated flow rate of 200 gpm. The expected frequency of this normal evolution is 212 flushes per year. Operators are in direct control of this evolution and procedural controls require that not more than 2000 gallons be used for each evolution. Gallons used are indicated in the Radwaste Control Room. This amount of flow and volume is well within the makeup capacity of the automatic valves and pumps that provide flow to the Fire Protection Water Storage Tank.

5. Seal water to Storm Drain Collector Basin Pumps Usage of seal water from the Fire Main to Storm Drain Basin pumps should be allowed because it is a small amount of flow. There are three pumps, and seal flow for all is not more than 40 gpm. Seal pressure is regulated to approximately 15 psig. Seal flow is used only when the pump is placed in service to lower Storm Drain Basin level. The expected frequency is dependent on rain fall.

Usage should be allowed because demand from these pumps is well below the capacity of the two fire pumps, which are sized to deliver 2000 gpm each. This is an insignificant amount when compared to the large volume of the Fire Protection Water Storage.

6. Temporary Cooling Water Supply to Service Air Compressor 1(2)D Usage of temporary cooling to Service Air Compressors should be allowed because this alignment is used typically once each refueling outage, and flow rates are well within the system capacity. Procedures require the affected unit to be in Mode 4 (i.e., Cold Shutdown) or Mode 5 (i.e., Refueling). Procedures direct that cooling flow to the air compressor be connected by a hose with pressure and flow regulated. Pressure and flow are estimated at 51 gpm and 44 psig. There is adequate margin in the capacity of the two fire pumps of 2000 gpm each and makeup capacity to the fire tank thru a 1.5-inch and 4-inch automatic make-up valve. Low tank level alarms are provided in the MCR. The Control Room Supervisor is in direct control of this procedure.

7. Refill of Standby Gas Treatment Drain Trough Usage of installed fire protection piping and valves should be infrequently allowed to fill the Standby Gas Treatment (SGT) drain trough because the flow and volume are insignificant when compared to the Fire Protection Tank volume.

The purpose of the trough is to ensure loop-seals can prevent by-pass leakage from the SGT filter compartments. Water level is checked regularly by plant operators and a small amount is added if needed to replenish evaporative loses. BSEP LAR Rev 2 Page L-6

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval Expected frequency is dependent on evaporation rate, normally less than twice per 24 hours. Total volume of each trough is approximately 60 gallons. To completely fill the trough from a normally closed 11/2-inch valve is insignificant compared to the volume of the Fire Protection Water Storage Tank and the design flow of the two fire pumps. Acceptance Criteria Evaluation: Nuclear Safety and Radiological Release Performance Criteria: The use of the fire protection water for plant evolutions other than fire protection is an infrequent, abnormal or emergency operational occurrence requiring Control Room Supervisor (CRS) direction and concurrence. The ability to isolate the non-fire protection flows ensures there is no impact on manual fire suppression efforts. Therefore, there is no impact on the nuclear safety performance criteria. The use of the fire protection water for plant evolutions other than fire protection involves fire protection water flow into existing plant systems. Leakage from these systems is not part of the fire protection system operation or firefighting evolutions and, as such, has no impact on the radiological release-performance criteria. Safety Margin and Defense-in-Depth: Since both the automatic and manual fire suppression functions are maintained, defense-in-depth is maintained. The methods, input parameters, and acceptance criteria used in this analysis were reviewed against that used for NFPA 805 Chapter 3 acceptance. The methods, input parameters, and acceptance criteria used to calculate flow requirements for the automatic and manual suppressions systems were not altered. Therefore, the safety margin inherent in the analysis for fire events has been preserved.

== Conclusion:==

NRC approval is requested for approval of the temporary use of the Fire Protection water supply, with the following restrictions: o Actions are controlled and described in approved Plant Procedures. o CRS approval is obtained. o Controls/communications are in place to ensure the non-fire protection system water demand can be secured immediately if a fire occurs. o Fire Protection Tank level shall be maintained with a minimum contained volume of 232,500 gallons (corresponding to a level of 24' 9 1/2"), and the Demineralized Water Tank, with a minimum contained volume of 90,000 gallons (corresponding to a level of 14' 0"). (Ref. OPLP-01.2) o Maintaining initial conditions to ensure no fire fighting operations are in progress. Page L-7 BSEP LAR Rev 2 LAR Rev 2 Page L-7

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval The engineering review of procedure guidance and controls in place for evolutions associated with use of Fire Protection Water determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3: o Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; o Maintains safety margins; and o Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). Approval Request 2 NFPA 805 Section 3.2.3(1) In accordance with 10 CFR 50.48(c)(2)(vii), "Performance-based methods," the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied. In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3: o Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; o Maintains safety margins; and o Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability). Duke Energy, BSEP requests formal approval of performance-based exception to the requirements in Chapter 3 of NFPA 805 as follows: NFPA 805, Section 3.2.3(1)

    "Proceduresshall be established for implementation of the fire protection program.

In addition to proceduresthat could be requiredby other sections of the standard, the proceduresto accomplish the following shall be established: Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program." Duke Energy, BSEP requests the ability to utilize performance-based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. Performance-based inspection, testing, and maintenance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1 006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection", Final Report, July 2003. BSEP LAR Rev 2 Page L-8

CP&L Aftachment L - NFPA 805 Chapter 3 Requirements for Approval Basis for Request: NFPA 805 Section 2.6, "Monitoring," requires that "A monitoring program shall be established to ensure that the availabilityand reliabilityof the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid." NFPA 805 Section 2.6.1, "Availability,Reliability, and PerformanceLevels," requires that "Acceptable levels of availability,reliability, and performance shall be established." NFPA 805 Section 2.6.2, "MonitoringAvailability, Reliability, and Performance," requires that "Methods to monitor availability,reliability,and performance shall be established. The methods shall consider the plant operating experience and industry operating experience." The scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the fire protection program have been established based on the previously approved Technical Specifications / License Controlled Documents and appropriate NFPA codes and standard. This request does not involve the use of the EPRI Technical Report TR-1006756 to establish the scope of those activities as that is determined by the required systems review identified in LAR Attachment C, "NEI 04-02 Table B Fire Area Transition." This request is specific to the use of EPRI Technical Report TR-1 006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program. As stated in EPRI Technical Report TR-1006756 Section 10.1, "The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability." This goal is consistent with the stated requirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained. The target tests, inspections, and maintenance will be those activities for the NFPA 805 required fire protection systems and features. The reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The failure criterion will be established based on the required fire protection systems and features credited functions and will ensure those functions are maintained. Data collection and analysis will follow the EPRI Technical Report TR-1 006756 document guidance. The failure probability will be determined based on EPRI Technical Report TR-1006756 guidance and a 95% confidence level will be utilized. The performance monitoring will be performed in conjunction with the Monitoring Program BSEP LAR Rev 2 Page L-9

CP&L Attachment L - NFPA 805 Chapter 3 Requirements for Approval required by NFPA 805 Section 2.6 and it will ensure site specific operating experience is considered in the monitoring process. The following is a flow chart that identifies the basic process that will be utilized. Program Framework IdentifyTarget Tests and Inspections Establish Reliability and Frequency pGals Set Failure Criteria Assess Licensing Impact and Other Constraints t i" E.: . i on.. Data Collection andEvaluation Establish Data Collection Guidelines Collect RRequired Surveillance Data Assemble Data in Spreadsheet or Database Analyze Data to Identy Failures

                              . * . .....
                                    . : . .......                      .                     . .  :. .  .

Reliability and Uncertainty Analysis 6Compute0Failure Probabilities:,

                                   'Compute Uncertainty.Limits Confirm That Reliability Supports Target Frequency
                                                                         ...... . '... ...          . .
                                                  .. . . .   . .         . . .. . . .. . .   .            .

Program Implementation........ Modify Program Documents Revise Surveillance Procedures. Conduct On~going .Performance Monitoring Refine and Modify F .requencies a.s:Aprrit EPRI TR-1 006756 - Figure 10-1 Flowchart for Performance-Based Surveillance Program BSEP LAR Rev 2 Page L-10

CP&L Affachment L - NFPA 805 Chapter 3 Requirements for Approval Duke Energy, BSEP does not intend to revise any fire protection surveillance, test or inspection frequencies until after transitioning to NFPA 805. Existing fire protection surveillance, test and inspection will remain consistent with applicable station, Insurer, and NFPA Code requirements. BSEP's intent is to obtain approval via the NFPA 805 Safety Evaluation to use EPRI Technical Report TR1006756 guideline in the future as opportunities arise. Duke Energy, BSEP reserves the ability to evaluate fire protection features with the intent of using the EPRI performance-based methods to provide evidence of equipment performance beyond that achievable under traditional prescriptive maintenance practices to ensure optimal use of resources while maintaining reliability. Nuclear Safety and Radiological Release Performance Criteria: Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to Nuclear Safety Performance Criteria by the use of the performance-based methods in EPRI Technical Report TR-1006756. The radiological release performance criteria are satisfied based on the determination of limiting radioactive release. Fire Protection Systems and Features may be credited as part of that evaluation. Use of performance-based test frequencies established per the EPRI Technical Report TR-1 006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited to meet the Radioactive Release performance criteria. Therefore, there is no adverse impact to Radioactive Release performance criteria. Safety Margin and Defense-in-Depth: Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited in the Fire Risk Evaluation safety margin discussions. In addition, the use of these methods in no way invalidates the inherent safety margins contained in the codes and standards used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analysis has been preserved. Page L-11 BSEP Rev 2 LAR Rev BSEP LAR 2 Page L-1 1

CP&L Aftachment L - NFPA 805 Chapter 3 Requirements for Approval The three echelons of defense-in-depth described in NFPA 805 Section 1.2 are:

1. to prevent fires from starting (combustible/hot work controls),

2. rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and

3. provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions).

Echelon 1 is not affected by the use of the EPRI Technical Report TR-1006756 methods. Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features credited for defense-in-depth are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2 and 3 for defense-in-depth.

== Conclusion:==

NRC approval is requested for use of the performance-based methods contained in the Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide", Final Report, July 2003 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. As described above, this approach is considered acceptable because it: o Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; o Maintains safety margins; and o Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). Page L-12 BSEP BSEP LAR Rev 2 LAR Rev 2 Page L-1 2

Enclosure 11 Revised NFPA 805 Transition Report, Attachment V, Fire PRA Quality

CP&L Affachment V - Fire PRA Quality CP&L Attachment V Fire PRA Quality

                                                    -

V. Fire PRA Quality 54 Pages Attached I Page V-I I BSEPLARRev2 I BSEP LAR Rev 2 Page V-1

CP&L Aftachment V - Fire PRA Quality The Fire PRA is adequate to support the NFPA 805 Licensing Basis. During the period of December 2011, the Boiling Water Reactor Owner's Group used the NEI guidelines to conduct a Peer Review of the Fire PRA based on the applicable requirements of the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) standard, ASME/ANS RA-Sa-2009, "Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications", as endorsed by Regulatory Guide 1.200, Revision 2. As assessed by the Peer Review team, the Fire PRA meets Capability Category II for most, but not all, Supporting Requirements. The Peer Review team noted a number of Facts and Observations (F&Os). Those F&Os that were characterized as Findings are listed in Table V-i, with both the disposition of the Finding and the status of the disposition. Where sufficient action has been taken for the Finding to be considered resolved, Table V-1 lists the Status as "Dispositioned;" otherwise the Status is listed as "Open." With the three exceptions listed below, the "Open" Findings are either documentation issues or ancillary statistical analyses. In general, while the future resolution of these Findings might facilitate better understanding of the model, the quantified risk results are not expected to be affected. For the following three exceptions, Table V-1 provides the basis for considering the Fire PRA to be sufficiently creditable for the NFPA 805 application:

1) Qualitative Evaluation of Equipment Susceptible to Smoke Damage (F&O 2-16),

2) Accounting for the State-of-Knowledge Correlation (F&O 4-18), and

3) Truncation Limits (F&O 1-36).

For the limited number of Supporting Requirements that were not assessed as meeting Capability Category II, Table V-1 also includes, as part of the disposition, a resolution of the Capability Category Classification and an evaluation of its impact on the NFPA 805 application. Finally, the Peer Review team identified the use of a split fraction for "Open"f'Closed" MCCs (F&O 4-1) to be an Unreviewed Analysis Method. However, as described in Table V-I, this approach was previously reviewed by the NRC as part of the Harris Nuclear Plant (HNP) NFPA 805 submittal and therefore, should not be considered to be an Unreviewed Analysis Method. l Page V-2 I I13SEP BSEPLARRev2 LAR Rev 2 Page V-2

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition ES-B2 Justification for Dispositioned Generic MSO scenario 2e appears to be inadequately The identified MSO scenarios were re-evaluated (CAT II) Exclusion of dispositioned. The scenario identified in NEI 00-01 is using deterministic and thermal hydraulic methods Generic MSOs a drain of the vessel, while the rough calculations for individual MSOs and combinations of MSOs. ES-D1 evaluate this as essentially (i.e., word in Attachment 3 The results of the re-evaluation concluded that the (CAT 1/11/111) of the component selection report) a depletion of the individual MSOs and combinations of the MSOs did PRM-B9 suppression pool. The loss is estimated as 200 gpm, not result in a failure of credited components, (CAT 1/11/111) which can be an issue either: a) long term for addition of new initiating events or a change in PRM-C1 inventory, or b) in combination with other small losses accident sequences. These MSOs remain as (See NEI 00-01 for guidance on combining MSOs). screened from inclusion in the FPRA model. The (CAT I/Il/Ill) documentation of the analysis of these MSOs has Generic Scenario 2d appears to also be a possible been updated in Attachment 3 of the component long-term issue (i.e., with multiple seal failures), or an selection calculation, BNP-PSA-085. issue in combination with other small losses. Scenario 1321-2c (i.e., Main steam drain line) includes an evaluation of flow size listed as 0.03 square inches based on a single flow path. However, multiple drain line openings are possible. (F&O 1-2) ES-A5 Spurious MSIV Dispositioned The MSIVs spurious operation appears to be modeled Given the MSIVs are normally open during power (CAT II) Operation as a failure of containment isolation under gate 1S1. operation, MSIV spurious opening or failure to close This spurious operation does not appear to be cannot be a fire-induced initiating event. ES-A6 modeled as either an initiating event or LOCA, or (CAT II) However, two new MSIV LOCA accident sequences showing to impact RCIC/HPCI operation. ES-B2 were created to model a fire-induced post-trip MSIV Most BWR FPRAs include MSIV failure to close or spurious opening or failure to close (MSO-B21-2b). (CAT II) spurious re-opening as a large or medium LOCA, These sequences do not credit HPCI or RCIC and ES-D1 given downstream opening of TBVs or other large include the loss of the condenser. (CAT 1/11/111) steam line valves. This has been documented in Section 3.3.1.4 and (F&O 1-6) Attachment 3 of the component selection calculation BNP-PSA-085. ES-Al Mapping Point Dispositioned The FPRA modeling does not include mapping of As a clarification, although the equipment selection (NOT MET) Estimate Initiating multiple point estimate initiating events to specific section of the FPRA standard requires the Events to Specific equipment. This includes the following lEs: Loss of identification of equipment whose failure could I BSEP LAR Rev 2 Page V-3

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition ES-A2 Equipment offsite power, Inadvertent opening of SRV (%lT-S), cause an automatic/manual trip or a mandated (CAT 1/11/111) Loss of DC Power (%lT DClAl, %lT-DClB2), Loss shutdown, neither the equipment selection section ES-A3 of Switchboard (%lTDClA, %1TOCiB), and loss nor the plant response model section requires the (NOT MET) of AC Bus (%lTE_El, %lTEE2). In essence, these mapping of specific equipment to any point estimate are treated as a plant transient (in this case, an MSIV initiating event. FQ-A2 closure event) followed by a subsequent failure of the (CAT 1/11/111) equipment. A detailed review of the fire induced initiating events was performed, with particular attention to those A sensitivity case was requested for the loss of DC initiating events identified by the Peer Review, and Power Al and loss of offsite power to determine the was documented in Section 3.3.1.4 and possible impact on the CCDP. The results show some Attachment 8 of the component selection calculation differences in the cutsets and the CCDP results, BNP-PSA-085. The review found that all initiating mainly due to actuation logic (applied under the IE events had been adequately addressed except for logic), restart logic, and failure of CRD. Overall, the fire induced LOOP. Logic for fire induced LOOP CCDP following the lEs is slightly higher than was added to the fault tree where appropriate. assuming the subsequent failure of the equipment. Inadvertent SRV opening was removed from IANAG005 which is present under IANAG 178 as Significant tracing was performed of the logic for each documented in Rev 2 of the change log, IE. In most cases, the IE logic was ORed with the Attachment 9 of BNP-PSA-085. Attachment 3 of the equipment logic. However, there were exceptions. For component selection calculation BNP-PSA-085 example, for the inadvertent SRV opening; gate documents MSOs that were evaluated as possible IANl G1 78 (HEADER A ISOLATED AND NOT initiators but determined not to be creditable. RECOVERED) included the IE but not the equipment logic for SRV opening. Another example: Gate #U1 3 RESOLUTION OF CAPABILITY CATEGORY (S2 LOCA OR SORV WITH ONE OR MORE SRVS CLASSIFICATION: FAILING TO RECLOSE) includes SRV logic above, but only for 2 or more SRVs. As a result, the single With the described changes incorporated, BSEP SRV opening for the IE is not included under this considers the risk results from the Fire PRA logic. model to be creditable for the NFPA 805 application and this finding to be sufficiently Review of LOOP logic indicated several locations resolved for both SR ES-Al and SR ES-A3 to be where consequential LOOP was not included; assessed as CAT 1/11/111 is MET. although the logic included in most cases other fire logic such as MSIV closure, or other assumed fire IEs. (F&O 1-8) I I BSEP LAR Rev 2 Page V-4

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition ES-Al Spurious Dispositioned Feedwater and HPCI overfeed is not included in the The applicable impact of Feedwater and HPCI (NOT MET) Operation of FPRA modeling for possible Fire-Induced Initiating overfeed, as initiator events, is already appropriately Feedwater and Events. modeled. ES-A4 HPCI as Initiating (CAT 1/11) (F&O 1-9) Because it does not degrade the ability of the plant Events to mitigate the resulting transient, Feedwater FQ-A2 overfeed (MSO-N21-2ai) was included in the FPRA (CAT 1/11/111) as an initiating event which is subsumed within the Turbine Trip initiator. This is consistent with the treatment of initiating events in the Internal Events model (BNP-PSA-032) and is supported by the results of the MSO Expert Panel review. Generically, NEI-00-01 does not list MSO-N21-2ai as applicable to BWR4s, noting that steam-driven feedwater pumps may not be a concern, and (upon review) the MSO Expert Panel concurred. Likewise consistent with the treatment of initiating events in the Internal Events model (BNP-PSA-032), the MSO Expert Panel did not consider a plant trip to be a creditable result of a spurious HPCI operation (MSO-E41-2u). However, the possible effect of spurious HPCI operation (MSO-E41-2u) on the ability of the plant to mitigate an otherwise initiated transient was considered. In particular, during a postulated spurious HPCI operation (MSO-E41-2u), the high RPV water level signal may not isolate the steam inlet valve, but Operating Experience suggests that the turbine would over speed on low quality steam and mechanically trip at some point prior to the RPV water level actually reaching the steam lines. Consequently, RPV water level is not anticipated to induce a concurrent RCIC failure. However, since the available Operating Experience does not specify the RPV water level at which the steam quality is assured to cause a turbine trip, the RPV water level is identified as a BSEP LAR Rev 2 Page V-5

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition source of uncertainty. Documentation to justify this position has been added to Section 3.1, Attachment 3, and Attachment 8 of the component selection calculation, BNP-PSA-085. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR ES-Al to be assessed as CAT 1/11/111 is MET. SY-B5 Power Dispositioned Instrumentation included in the FPRA that affects The power supplies for the instrumentation credited (CAT 1/11/111) Supplies of HFES are listed in Calculation BNP-PSA-084, for operator actions, as identified in BNP-PSA-084, Instrumentation Revision 1, attachment 4. This attachment provides a have been added to the FPRA. The revision is SY-A6 Credited for comprehensive list of instruments affecting each of documented in the component selection calculation (NOT MET) Operator Actions the modeled HEPs in the PRA. (BNP-PSA-085) model change log. Power supplies SY-B9 were already included in the component selection to However, the power supplies for the instrumentation (CAT 1/11/111) support other modeled equipment. added to the FPRA model is not included in the FPRA ES-Cl logic RESOLUTION OF CAPABILITY CATEGORY (CAT 1/11/111) CLASSIFICATION: (F&O 1-10) PRM-B9 With the described changes incorporated, BSEP (CAT 1/11/111) considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR SY-A6 to be assessed as CAT 1/11/111 is MET. I Page V-6 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-6

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition PRM-B4 Additional Dispositioned Change Package BNP-0122 includes a list of plant The BNP trip assessment in Change Package (CAT I/Il/Ill) Impacts on Plant areas, and an evaluation of a possible plant trip for BNP-0122 (Attachment 10 of BNP-PSA-080 R1) Trip Likelihood each area. The categories include near certainty plant was updated using additional insight of targets in trip (1.0), reduced likelihood trip (0.1) and plant trip each fire compartment/zone. Both targets identified not likely (0.01). for the safe shut down and the fire probabilistic assessment were considered. In discussions with the engineer who developed this list, the assessment was based on judgment, which including some consideration for the likelihood of the fire, some consideration of the possible damage of a fire, and the equipment in the area. However, the judgment did not include a review of cables and equipment impacted in each area. The results show that areas impacting safety busses (i.e., which would result in a likely rapid plant shutdown), are estimated to shut down the plant 10% of the time, while impacts from spurious operation (e.g., SRV openings, MSIV closures) are not accounted for. Additionally, the base assumption of all fires causing a plant trip, loss of feedwater, loss of condenser vacuum, and MSIV closure (e.g., no cable tracing for these initiating events) is not applied. (F&O 1-14) FSS-A1 ZOI Approach for Dispositioned The transient ZOI approach was based on the 75th More accurately, the transient ZOI was based on a (NOT MET) Scenarios fire versus the 98% fire. As a result, the transient 143 kw 98% HRR in NED-M/MECH-1 006, Generic Involving scenarios were impacted as follows: Fire Modeling Treatments, rather than the 317 kW Transient Ignition 98% HRR in NUREG/CR-6850.

1) Scenarios were not identified in areas where the Sources cable trays were above 6 feet, but below the zone 1) Except for the turbine building, the ZOI for the of influence for a 317 kW fire (i.e., height depends lower HRR was retained for transient ignition on location). sources. This was based on existing and/or

2) Area for the ZOI was limited. For example, in the planned administrative controls and is supported by plant experience and by risk insights from a cable room, the area for each transient scenario was typically 3' x 3', versus a longer area which bounding sensitivity study. Primarily to preclude the imposition of more restrictive administrative may impact a particular cable tray. Again for this I Page V-7 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-7

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition area, several cable tray runs are 30' or longer, controls, the turbine building was re-examined for where the area assumed for a larger ZOI would a transient fire representing a 317 kW 98% HRR, be something like 30' x 7.' as documented in Attachment 16 of Revision 1 of BNP-PSA-086, and the results were incorporated

3) Areas, such as the Battery Rooms have no into the Fire PRA. For other parts of the plant, the identified transient scenarios.

planned use (post-transition to NFPA 805) of (F&O 1-19) FIR-NGGC-0009, is credited with limiting the placement of transient combustibles and ignition sources near equipment and cables unless a specific evaluation is performed using a 317 kW 98% HRR. In anticipation of possible future needs, other parts of the plant have been re-evaluated for a transient fire representing a 317 kW 98% HRR, but the results have not been incorporated into the Fire PRA. As documented in Attachment 25 of Revision 1 of BNP-PSA-086 (i.e., Change Package BNP-0220), a sample of plant transient combustible walkdowns (i.e., recorded over the last two years) was reviewed to determine a ZOI for a reasonably realistic and bounding transient ignition source. In addition, the risk associated with using a larger ZOI for a transient in a particular area was approximated in a bounding sensitivity study as a hot gas layer. Change Package BNP-0227 documents the source and target walkdowns that were conducted for those parts of the plant that experience larger transient combustibles and exhibit greater risk. These results may be incorporated into the Fire PRA as needed to support the plant and as controlled by FIR-NGGC-0009.

2) The floor area applied for each transient scenario is based on the identified target set.

The minimum applied transient foot print is 3'x3'.

3) All plant areas were assigned transient ignition I Page V-8 I BSEPLARRev2 I BSEP LAR Rev 2 Page V-8

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition source(s). If the transient ignition source did not damage any significant targets, no risk increase would be recorded from that potential fire source. This was the case for the Battery Room. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: The lowering of the HRR for transients, based on fire modeling insights and stricter controls on transient combustibles in some areas, follows an approach piloted at HNP. In Table 3.4-6 of Attachment C6 of the Safety Evaluation for the HNP license amendment (ML101130535), the NRC staff concluded that the approach was a reasonable and acceptable exception to using the bounding values from NUREG/CR-6850. With the described changes incorporated, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-A1 to be assessed as CAT 1/11/111 is MET. FSS-A1 Transient ZOI for Dispositioned Transient scenarios are identified using a ZOI Consistent with the guidance in H.2 of (NOT MET) Other Than Cable assuming cable damage only. No damage to NUREG/CR-6850, all of the ZOIs are based on Damage equipment appears to be assumed for any area. cable damage. It would be very conservative to assume equipment damage based on the same For example, a transient fire in the battery room was ZOI. In most cases the equipment is shielded to not developed where the transient damages or ignites some degree by a steel enclosure, such that internal the batteries, which is near the floor. Another example damage would be minimal from an external source. is there are no scenarios located between 1CB and Also, exclusion zones exist to limit placement of 1CA, where damage to both cabinets may occur. unattended transient ignition sources next to MCCs / (F&O 1-20) energized equipment (ref. OFPP-014, 5.22 and FIR-NGGC-0009, 9.1.12). I Page V-9 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-9

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-A1 to be assessed as CAT 1/11/111 is MET. HR-El Identification of Dispositioned No new actions were identified in response for the Following the Peer Review, a detailed HRA was (NOT MET) New Operator Fire. Discussions with BSEP operators, the ASSD developed to provide a more realistic evaluation of a Actions for Fire procedures will be used for shutdown given a fire and remote shutdown following control room HR-E2 (NOT MET) damage to ASSD equipment. For example, ASSD-01 abandonment. A review of ASSD-01 and ASSD-02 will call for shutdown outside of the control room in identified key operator recovery actions and related HR-E3 ASSD-02. For other areas, there are specific ASSD system interfaces. A proper understanding of (CAT Il/111) procedures. ASSD-05 was reviewed for fire in Unit 1 system operation within the context of a fire scenario Reactor Building North. This procedure includes was obtained during focused talk-throughs and HR-E4 specific recovery actions and manual actions, operator interviews. The results of the HRA, (CAT Il/111) including for example operation of the SRVs from the including the operator interviews, are documented in FSS-B1 RSP. Neither the control room evacuation actions nor Attachment 10 of BNP-PSA-084, Revision 2. (CAT 1/11/111) the local manual actions were identified or reviewed as a part of the fire PRA. As a result, the FPRA Possible conservatisms associated with not HRA-A2 results are conservative. For example, the top cutset modeling other ASSD actions are not considered to (NOT MET) be significant. for the cable room could be recovered using a control HRA-C1 room evacuation action. RESOLUTION OF CAPABILITY CATEGORY (CAT II) CLASSIFICATION: (F&O 1-24) HRA-D1 With the described changes incorporated, BSEP (NOT MET) considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR HR-El, SR HR-E2, and SR HRA-A2 each to be assessed as CAT 1/11/111 is MET and for SR HRA-D1 to be assessed as CAT II is MET. I Page V-lO I I BSEP BSEPLARRev2 LAR Rev 2 Page V-1 0

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition HR-G1 HEP for Control Dispositioned The control room abandonment HEP for habitability A detailed Human Event Probability (HEP) has been (CAT I) Room scenarios uses a CCDP of 0.1 and CLERP of 0.01, developed for the control room abandonment Abandonment without detailed analysis or support. These values scenario. An evaluation of the various key operator FSS-B2 may be conservative or non-conservative, depending actions contained in the abandonment procedures (CAT II) on the scenario (including equipment damage) and was performed using the CBDTMETHERP HRA-Cl timing. methodology contained in the HRA calculator. The (CAT II) No detailed timing, feasibility, review of training, evaluation uses Safe Shutdown timing studies and review of procedures, or detailed task analysis was feasibility analysis. The specific training and documented in the FPRA. frequency of training was evaluated as well as a detailed review of the procedure. Significant (F&O 1-26) equipment failures were also considered in the determination of the CCDP. The HEPs resulting from the HRA calculator evaluations were then placed into an event tree with supporting top logic to determine an overall HEP. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With the described changes incorporated, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR HR-G1 to be assessed as CAT 11is MET. I Page V-lI I IBSEPLARRev2 BSEP LAR Rev 2 Page V-11I

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-Al Propagation of Dispositioned MCC fire scenarios do not include propagation from When an open MCC fire was modeled, it is (NOT MET) MCC Fire one MCC stack to another. NUREG/CR-6850 conservatively assumed that the entire MCC is failed includes a propagation model, where propagation is and all targets within the ZOI of the MCC are treated assumed following a 10-15 minute delay (i.e., as failed by the fire. When the cabinet remains depending on the opening). closed, the fire is assumed to remain confined to a single stack. Insights are based on pilot plant The BWROG methods (i.e., not approved) include a evaluation of Fire Induced Flow Within a Motor probability of propagate and an approach for limiting Control Center (i.e., HNP-M/MECH-1 207). the number of cabinets considered in propagation, Additional reviews of BNP MCCs were performed to and an approach for determining the HRR. determine the applicability of conclusions of the (F&O 1-30) HNP calculation to MCCs at BSEP. This review concluded that the BSEP MCCs are of similar construction to those at HNP therefore the assumption that in-cabinet fires remain confined to a single stack applies also to the MCCs at BSEP. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-A1 to be assessed as CAT 1/11/111 is MET. FSS-Cl Fire Scenario Dispositioned

• Location factors (e.g., wall effects) were not Location factors were not included in the original (CAT I) Development included in HRR calculations for transients. walkdowns because they were determined to add
• Review of the documentation shows that ceiling little value considering the very large uncertainties associated with modeling transient fires, including jet treatment was not performed.

size of the transient taken into the room, the HRR of The 75th and 98th percentile HRR assigned for that specific transient package, and the pumps (electrical fire) are from Case # 7 for approximation for achieving an increase in ZOI due motors, BIN 14 (69 kW) in lieu of from Case # 6 to wall effects. New walkdowns were performed for for pumps, BIN 21 (69 and 211 kW). transient sources in the Turbine Buildings, where the largest uncertainties exist over transient package (F&O 1-32) size which would affect the size of wall effects. The new walkdowns increased the HRR and accounted I Page V-12 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-1 2

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition for wall effects. These have been incorporated into the Fire Scenario Data calculation (i.e., BNP-PSA-086). Use of wall effects in other areas has been added as an uncertainty in the calculation. The identification of targets is based on a ZOI determined from the source HRR using accepted and approved methods. Where secondary fire growth is expected, The ZOI treatment is conservatively extended to the ceiling. The treatment of ceiling jets would only be addressed when more detailed fire modeling is applied. This does not typically apply to transient sources since the overall transient analysis is based on virtual sources and does not contain the specific inputs that would be needed to justify the applicability of detailed analysis. While NUREG/CR-6850 recommends the use of 211 kW for pumps, there are two footnotes indicating that there is no experimental evidence for the HRRs which are conservatively based on electrical cabinet fires. Examinations of motors driving pumps versus other motors does not reveal significant differences that would lead one to conclude the motor fires would be more severe for pumps, unless oil was involved. Since the oil fires are handled as a separate scenario already, it does not need to be included in the motor fire HRR. Examination of motors, as presented in FPIP-0150, Ignition Source Characterization and Fire-Related Assumptions, indicate no significant variance for either application. Consideration of fire potential and associated HRR's provides qualitative similarities to that of a dry transformer (i.e., NUREG/CR-6850, Table 11-1), which also uses a HRR value of 69kW. Further, if there was a significant electrical cabinet associated with the pump motor, it was also given its own I I BSEP LAR Rev 2 Page V-1 3

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition scenario using the electrical cabinet HRR. Therefore, it is deemed within the bounds of realism to use the motor specific HRR for pump motors. This deviation from NUREG/CR-6850 has been added as a source of uncertainty to BNP-PSA-086. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With the incorporation of the described limited changes, BSEP considers the bounding risk results from the Fire PRA to be acceptable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-Cl to be assessed as CAT II is MET. FSS-G2 Multicompartment Dispositioned A screening value for rated barrier probability of 1E-2 The BSEP fire quantification calculation has been (CAT 1/11/111) Screening Value was applied. This may not be bounding depending on revised, and the screening of HGL Multi the features of the barrier (i.e., doors, penetrations, Compartment Analysis has been performed in dampers). accordance with NUREG/CR-6850. The screening (F&O 1-34) value of 0.1 was used on the exposing compartment to screen out compartments from the MCA analysis. The results of the revised Multi-Compartment Analysis are documented in the quantification calculation BNP-PSA-080. I Page V-14 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-14

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition QU-B2 Truncation Limits Dispositioned Truncation in the CDF and LERF was varied, based The truncation approach has been changed in Rev 1 (NOT MET) upon the CCDP/CLERP. For example, CCDP of 1.0 of the quantification calculation (i.e., BNP-PSA-080) uses a truncation of 1.0, while a CCDP of 1E-03 uses in response to this F&O. Scenarios are now run at QU-F2 a truncation of 1E-07. Overall, the process using the an effective truncation of 1E-09/yr for CDF and (CAT 1/11/111) one run results in difficultly running FRANC at a very 1E-1 0/yr for LERF, which is more than four orders of QU-B3 low cutoff. magnitude below the resulting CDF and LERF plant (NOT MET) totals. A review of the truncation levels was performed. FQ-B1 Hundreds of the sequences have truncation within a RESOLUTION OF CAPABILITY CATEGORY (CAT 1/11/111) factor of 100 or less of the CCDP. Several of these CLASSIFICATION: sequences were re-run, and the new CDFs were FQ-F1 Since the process for establishing truncation limits compared to the original CDFs. Changes in the (NOT MET) does not demonstrate that the overall model results vary from about 5% to as much as 25%. Many of the sequences affected are in the top 25 fire results converge, SR QU-B3 will continue to be sequences. assessed as NOT MET. However, the very low effective truncation (i.e., relative to the resulting Additionally, a large number of scenarios are listed CDF and LERF plant totals) provides reasonable with zero CCDP. When these were re-run with lower assurance that no significant accident sequence truncation values, cutsets were generated. This can was inadvertently eliminated. be important for scenarios with higher ignition frequencies. With the incorporation of the described limited changes, BSEP considers the risk results from (F&O 1-36) the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for both SR QU-B2 and SR FQ-F1 to be assessed as CAT 1/11/111 is MET. I Page V-15 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-1 5

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-G2 LERF Uncertainty Open A quantitative evaluation of LERF uncertainty was not LERF numerical uncertainty is deferred for future (NOT MET) included in the final results. The uncertainty analysis. The uncertainty evaluation does not affect quantification was performed for CDF results only. the quantified baseline Fire Risk. Most of the LE-F3 Assumptions and key areas of uncertainty did not uncertainties are in the conservative direction based (NOT MET) include discussion of LERF, other than the use of a on the typical modeling practice to initially use UNC-A1 simplified LERF value for control room abandonment. screening/bounding values for inputs. (NOT MET) (F&O 1-38) RESOLUTION OF CAPABILITY CATEGORY FQ-E1 CLASSIFICATION: (NOT MET) Although no change has yet been made that FQ-F1 would improve the Capability Category (NOT MET) assessments, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because a quantitative evaluation of LERF uncertainty will not change the quantified risk metrics. CS-Al Fire PRA Cable Dispositioned The BSEP FPRA roadmap indicates that the In the process at BSEP, Fire Protection/NSCA (CAT 1/11/111) Selection methodology to identify additional cables uses the develops and maintains the cable selection and Notebook same process for PRA circuit analysis as for the circuit analysis data. These data are then CS-A3 deterministic Safe Shutdown circuit analysis. referenced as inputs to the Component Selection (CAT 1/11/111) Reference FIR-NGGC-0101. and Quantification FPRA calculations. This process CS-Cl and associated results are easily reviewable, has However, there is no separate notebook for Fire PRA (NOT MET) been peer reviewed multiple times for our other sites Cable Selection to discuss the processes, inputs, and and found to be acceptable. There is no results. requirement to have a separate PRA notebook. (F&O 2-2) RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR CS-Cl to be assessed as CAT I/Il/Ill is MET. I Page V-16 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-1 6

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition QLS-A1 Qualitative Dispositioned The QLS screening criteria may not have been Revision 2 of BNP-PSA-083 removes FC261 (CAT 1/11/111) Screening of applied appropriately. BNP-PSA-083 Rev. 1 Section (DUCTBANK) from qualitative screening and retains DUCTBANK 3.3 documented the screening criteria as in it for quantitative analysis. QLS-A2 NUREG/CR-6850. Alternate screening criteria was (CAT 1/11/111) Raceway target information, cable loadings, and used to screen several analysis units. These criteria QLS-A3 were based on the judged low risk significance of the floor areas for the manholes in the DUCTBANK (CAT 1/11/111) unit in question. were collected, MOS factors were assigned, and transient ignition frequencies were determined. QLS-A4 Moreover, FC261 (DUCTBANK) was screened out (CAT 1/11/111) based on no equipment while the QLS screening With the exception of FC261 (DUCTBANK), no criteria need to rule out both equipment and cables. physical analysis unit was qualitatively screened QLS-B3 based on the use of alternate screening criteria. DUCTBANK will contain a large number of cables and (CAT 1/11/111) low risk contribution is not expected. However, FC295 and FC345 (i.e., Drywell/Torus, for Unit 1 and Unit 2, respectively) were not retained for The BSEP team responded as follows: quantitative analysis because no ignition frequency

                                                'FC261 (DUCTBANK) is not a typical fire                    was assigned to the Drywell/Torus based on the Technical Specifications requirements for an inert compartment (i.e.,'... a well-defined enclosed atmosphere during power operations. This room...'). As described in Attachment 3, FC261 (DUCTBANK) is a network of underground conduit in          treatment is consistent with both the Fire Hazard pre-cast concrete cable trenches. Rather than              Analysis in the (U)FSAR and the Safe Shutdown Analysis, in which no fire is postulated in or subsuming FC261 (DUCTBANK) into FC263 (with analyzed for the Drywell.

certain 'yard' locations), FC261 (DUCTBANK) was separately identified during plant partitioning to promote clarity in communication with legacy plant fire protection programs. Because of its design, no transient fire was postulated for FC261 (DUCTBANK). As described in Attachment 3, no equipment is located in the FC261 (DUCTBANK). And as stated in Section 3.4.2, all cables at BSEP are considered qualified self-extinguishing and non-propagating. With no creditable ignition source, there is no fire risk. Therefore, it was considered appropriate to qualitatively screen FC261 (DUCTBANK) consistent with the stated intent of the general task objective described in Section 4.3.1 of NUREG/CR-6850.' I Page V-17 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-1 7

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition However, it is not expected to have absolute zero fire ignition frequencies in the ductbanks since these enclosed areas could be open for maintenance during outages and transients could be left there unnoticed, even the transient materials from plant startup. On the other hand, 100% qualified self-extinguishing and non-propagating cables may not be realistic. Cables used for lighting and other not modeled system functions may exist in the ductbanks, which may not be qualified. (F&O 2-3) I Page V-18 I BSEPLARRev2 I BSEP LAR Rev 2 Page V-1 8

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition SY-Cl Updated System Open The system notebooks (i.e., calculation The update of the system notebook (i.e., (NOT MET) Notebooks BNP-PSA-062) have been updated for MOR1 1 from BNP-PSA-062) with the relevant fire information which the FPRA was subsequently developed. The occurs as part of the normal PSA model update SY-A2 system notebooks (i.e., calculation BNP-PSA-062) will process used for the entire BSEP PSA. (NOT MET) be further updated to incorporate fire-specific changes RESOLUTION OF CAPABILITY CATEGORY SY-C2 to the model. CLASSIFICATION: (NOT MET) However, the system analysis supporting requirements included in SY-A2, A3, A4, A6, C1 and Although no change has yet been made that SY-A3 C2 have been determined to be not met with the would improve the Capability Category (NOT MET) assessments, BSEP considers the risk results current documentation, which was typically performed SY-A4 from the Fire PRA to be creditable for the by updating the system notebooks to reflect all fire-(CAT I) NFPA 805 application because adding the related changes. additional pertinent fire model documentation to SY-A6 An example of information not included from SY-A2 the system notebooks will not change the risk (NOT MET) includes: insights and metrics. PRM-B9 COLLECT pertinent information to ensure that the (CAT 1/11/111) systems analysis appropriately reflects the as-built PRM-C1 and as-operated systems. Examples of such (CAT 1/11/111) information include system P&lDs, one-line diagrams, instrumentation and control drawings, spatial layout drawings, system operating procedures, abnormal operating procedures, emergency procedures, success criteria calculations, the final or updated SAR, technical specifications, training information, system descriptions and related design documents, actual system operating experience, and interviews with system engineers and operators. See other referenced SRs for other information not included in the FPRA documentation. (F&O 2-8) I Page V-19 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-1 9

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition SY-C3 Updated Open The system notebooks (i.e., calculation The update of the assumptions and sources of (NOT MET) Documentation of BNP-PSA-062) have been updated for MOR1 1 from uncertainty related to fire-specific changes to system Model which the FPRA was subsequently developed. The modeling are captured and updated as part of the PRM-C1 Uncertainty and system notebooks (i.e., calculation BNP-PSA-062) will normal update process for the BSEP PSA and (CAT 1/11/111) Related be further updated to incorporate fire-specific changes would be added to the BNP-PSA-075 calculation. Assumptions to the model. However, the sources of model RESOLUTION OF CAPABILITY CATEGORY uncertainty and related assumptions are not CLASSIFICATION: documented. Although no change has yet been made that (F&O 2-9) would improve the Capability Category assessment, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because updating the documentation of sources of model uncertainty and related assumptions will not change the quantified risk metrics. QU-E4 Uncertainty Dispositioned PRA items that were assumed failed for the The requested sensitivity is on items considered (CAT 1/11/111) Analysis of component selection are listed in BNP-PSA-085 always failed in the Fire PRA. This treatment Components Rev. 1 Section 4 and Table 4. represents a conservatism in the calculated Fire UNC-Al Assumed Failed CDF. Of these, the largest effect is likely the (NOT MET) This treatment is similar to treatment of unknown assumption of loss of feedwater for each scenario. locations for equipment that do not have cable-routing PRM-B1 0 completed. RESOLUTION OF CAPABILITY CATEGORY (CAT 1/11/111) CLASSIFICATION: Sensitivity studies should be performed to investigate the risk importance of these failed systems/functions. With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for (F&O 2-10) the NFPA 805 application and this finding to be sufficiently resolved for SR UNC-A1 to be assessed as CAT 1/11/111 is MET. I Page V-20 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-20

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-D7 Outlier Review of Dispositioned BNP-PSA-080 Section 4.5.3, Non-Suppression Section 3.2.3.3 of BNP-PSA-083 documents (CAT I) Fire Detection Probability, documents the methods used for consideration of the applicability of using generic and Suppression calculation of non-suppression probabilities. Generic non-suppression data based on an outlier review of System NSP and unavailability are applied from plant fire bridge experience. Currently, system Unavailability NUREG/CR-6850. No outlier review is performed, and performance is monitored and maintained at a high no plant specific data are used to update the level as part of the System Health Reporting and unavailabilities. System Notebook processes. Outlier behavior with respect to system availability would be evident to the (F&O 2-14) system engineer and plant management through the health data (available for the previous 12 months), which indicates overall Excellent (Green) performance. Post-transition, the assessment of system performance is part of the NFPA 805 Monitoring Program, as described in procedure FIR-NGGC-O1 30. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: Using plant-specific information to quantify total unavailability factors is a CAT Ill requirement and was not done. With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-D7 to be assessed as CAT II is MET. I Page V-21 I I BSEP BSEPLARRev2 LAR Rev 2 Page V-21

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-D8 Specific Area Dispositioned The note of SR FSS-D8-1 states: Fire detection or Accommodation of area specific features and (CAT 1/11/111) Features and suppression system effectiveness depends on, at a scenarios is assured for fire suppression and Suitability of minimum, the following: 1) system design complies detection system through correct application of fire Installed System with applicable codes and standards, and current fire protection design standards such as NFPA 13 protection engineering practice, 2) the time available Standard for the Installation of Sprinkler Systems, to suppress the fire prior to target damage, 3) specific and NFPA 72 National Fire Alarm Code. In each features of physical analysis unit and fire scenario case, careful selection of occupancy classification under analysis (e.g., pocketing effects, blockages that and hazard identification is applied. This ensures might impact plume behaviors or the "visibility" of the that physical features and the fire sources contained fire to detection and suppression systems, and in a given area are properly protected to achieve the suppression system coverage), and 4) suitability of desired performance results. Ceiling the installed system given the nature of the fire source configurations, blockage of agent application by being analyzed. design features and adequate coverage for the hazards present are a direct function of code In light of B3SEP fire scenarios, above item 1 should compliance. Code compliance is further assured by be considered met although not evident in detailed evaluation in the NFPA 805 Transition documentation. Timing (i.e., item 2) is considered in report Table B-i, through the use of or reference to detailed NSP calculations were carried out in the Code Compliance Calculations such as OFP-1038, spreadsheet files BNP_-EVALUlCDF.xls, Rev. 1, Code Compliance Evaluation NFPA 13 BNP_-EVALUlLERF.xls, BNiPEVAL__U2_CDF.xls, (Reactor Building), 1976 and 1983 Ed. or OFP-1031, and BNPEVALU2_LERF.xls. Rev. 0, Code Compliance Evaluation NFPA 72E, However, above items 3 & 4 were not addressed. 1984 Ed. for BSEP. (F&O 2-15) I Page V-22 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-22

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-D9 Qualitative Open BNP-PSA-086, Section 10.0, states that fires resulting The impact of smoke damage has not been (CAT I) Evaluation of in significant smoke production could cause additional evaluated in detail, because there is no approved Equipment damage beyond the heat based zone of influence method for doing so. However, the incremental Susceptible to target sets collected. However, targets that are impact of smoke effects is expected to be somewhat Smoke Damage susceptible to smoke damage have not been mitigated by the fire affects of a hot gas layer, which identified and are currently not evaluated in this has already been evaluated. In general, the calculation. Therefore, this SR is considered not met. combination of fire scenarios and locations that (F&O 2-16) might favor the production and concentration of sufficient smoke, to damage additional equipment beyond the heat-based zone of influence, is expected to favor also the formation of a hot gas layer which would damage the cables to that same equipment. Consequently, the target set for smoke damage would likely be similar to the target set for a hot gas layer, which is already evaluated. This has been included as a source of uncertainty in the quantification calculation, BNP-PSA-080. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: Although no change has yet been made that would improve the Capability Category assessment, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because conditions favoring smoke damage are considered generally similar to those favoring the formation of the more bounding hot gas layer. I Page V-23 I BSEPLARRev2 BSEP LAR Rev 2 Page V-23

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FQ-A1 Excessive Dispositioned A review of FRANC model files showed that some The FPRA database query (CAT 1/11/111) Mapping of HGL scenarios (i.e., whole room burnout) have less q__,SourceBE_2a(source), which identifies the failed Components for affected components than some individual scenarios components for the individual scenarios, was Individual in the same fire compartment (FC) modeling a single modified to use the same FSSPMD mapping table Scenarios ignition source and targets in its ZOs. (T._RoutingFireZone) that is used for generating the HGL component failures (i.e., Reference For example, in the Unit 1 CDF FRANC model, BNP-PSA-080 Rev 1). FC212 scenario BHGL has 64 affected components while scenarios FC212o4612 B75 and B98 have 112 affected components. On the other hand, some other scenarios have significantly more affected components in HGL scenarios than individual scenarios in the same FC. Discussion with BSEP PRA team indicated that different mapping tables have been used for HGL scenarios and individual ignition source scenarios. Conservatism may exist in the generation of mapping tables for individual scenarios. However, non-conservatism could exist for the HGL scenarios if the different mapping tables do not cover all the cablesI equipment that are affected by the fire-induced failures. (F&O 2-19) I Page V-24 I I BSEP BSEPLARRev2 LAR Rev 2 Page V-24

CP&L Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FQ-A1 Crediting Spray Dispositioned FC238_5010 and similar fire scenarios are expected Metal water spray shields are provided over several (CAT 1/11/111) Shields in the to have significant SBO contributions. A review of fire initiators in the Diesel Generator Basement. Diesel Generator mapping table and excluded events and altered event Specifically these metal shields are installed over Basement tables did not show the failure of DG Breaker spurious the EDG Excitation Voltage Source PT & Reactor failures excluded, which is also evident in FRANC and EDG Excitation Current XFMR Phase A, B, & C affect components. However, it is noted that the IGN as documented in drawing F-1319. The spray is set to 0. shields are BSEP plant configuration and are maintained via controlled drawings. The shields are BNP PRA team responded that: designed to prevent water spray impingement onto

                                                'During review of cutsets following preliminary             the transformers described above and, per quantification, several scenarios were identified as        controlled drawing examination and plant walkdown, significant contributors to plant risk. Review of these     they also provide a non-combustible barrier to the scenarios identified significant conservatisms in the       development and passage of a damaging fire plume initial data inputs that were causing unrealistic risk      above these transformers. Based on walkdowns, results. As part of this review, it was identified that the the construction of these shields is sufficient to fire size for sources 5010 through 5017 were initially      prevent direct passage of a damaging fire plume to characterized as 211 kW fires when a more detailed          targets located directly above the protected examination of the equipment showed that the                transformers.

sources should be characterized as 69 kW fires. In The primary concern with a fire in the subject addition, a shield above the sources was identified. transformers (i.e., sources 5010, 5011, 5012, 5013 Consideration of either of these two facts would result 5014, 5015, 5016 and 5017) is development of a fire in consequences for the fires that would be much less plume that would impact cable trays routed above severe than the initial walkdown information indicated. the spray shield. The design of the spray shield is This information is documented in change package such that the plume would be forced to follow a BNP-0182 and BNP-0176 (see the BNP-0176 change circuitous path prior to impingement on the target package directory of BNP-PSA-080 calc for pictures cable tray. The worst case fire expected to develop of these sources). Because the quantification process in the fire initiators would be a 69 kW fire based on was nearly complete, explicitly incorporating the the 98% HRR for dry-type transformers, Ref. information from the change packages into all input NUREG/CR-6850. calculations would have resulted in a significant administrative burden to revise the calculations. All of the cables located in the Diesel Generator Therefore, to simulate the correct effects within the Basement are IEEE 383 qualified; therefore, their quantification calculation, the scenarios were damage temperature is 625OF and damaging heat assumed to be equivalent to the first target tray flux is 1lkW/m2. The target cable trays are located having a solid bottom as per BNP- 0176 and the above the EDG transformer spray shields therefore scenario event frequency was set to zero for damaging temperatures must be exceeded at the scenarios FC238_5010 B75 and FC238_5010 B98. It spillage points of the spray shield to be deemed I Page V-25 I BSEPLARRev2 BSEP LAR Rev 2 Page V-25

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition is assumed that cable trays with solid bottoms will capable of damaging the cable trays or a damaging prevent damage to cables for ignition sources with radiant heat flux radiated from the spray shield. HRR 69 kW or less based on the discussion provided Based on review of the spray shield design and in section Q.2.2 of NUREG/CR-6850.' plant walkdown of the initiator/target configurations, it is judged that the spray shields installed above A test fire model run with FDS was also constructed these transformers will prevent thermal damage to to demonstrate the adequacy of above engineering the target cable trays; thus, damage resulting from a judgment. As a result, the technical basis supporting transformer fire need not be postulated. the treatment of the identified scenarios is considered acceptable. However, the following issues should be Section 8.5.3 of BNP-PSA-080 Revision 2 provides addressed: risk insights associated with removal or failure of these metal shields.

1. The documentation in BNP-0176 should be enhanced to include engineering judgment as Continued maintenance of these spray shields is discussed above instead of a simple assumption ensured by plant documentation and credit for these that the metal cover above the cabinet is sufficient spray shields as a radiant/plume shield for raceways in preventing fire damage to targets above the located above the EDG transformers is documented cover. in the fire PRA calculation.

2. The BNP team stated that the metal cover is part of the design basis. This fact should be verified and documented in fire PRA.

3. The potential failure of the metal cover should be addressed. May need to credit the surveillance /

inspection / maintenance program to ensure the integrity of this metal cover.

4. Perform sensitivity study or include the failure probability of this metal cover to generate risk insights associated with the assumption associated with this metal cover.

5. Revise the fire PRA model to not set ignition frequency to 0 but remove the impacted targets.

(F&O 2-20) CF-Al Conditional Dispositioned BNP-PSA-080 Section 4.3.4, Fire Induced Spurious The listed non-instrument spurious cable failures (CAT I) Failure Event Probabilities, document the methods used for were analyzed, and probabilities were included in I I BSEP LAR Rev 2 Page V-26

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition CF-B1 Probabilities for conditional failure probabilities for fire-induced circuit the Fire PRA. Conditional failure probabilities were (CAT 1/11/111) Fire-Induced failures. assigned to the most risk significant contributors, Circuit Failures causing them to become less risk significant and Circuit Analysis was performed in change package BNP-01 37 to determine the probability of a spurious allowing these less risk significant contributors to appear relatively more risk significant. More could operation for various cables. have been done, but the iterative process stopped Risk significant contributors were not identified when satisfactory results were obtained. (quantification was complete later in the process) and In many of the identified cases, failures are in utilized thus cannot met the capability category CC-Il. instrumentation, and probability analysis methods For example, the Unit 1 CDF importance results are not available, and no testing has been done to include the following spurious events for which determine the failure probabilities. Division of failure conditional probabilities have not been developed: mode based on conditional probability analysis would only serve to add additional uncertainty to the HPC1PPS-SA-N12A T, PRESSURE SWITCH failures. E41-NO12A SPURIOUSLY ACTUATES The current analysis is conservative in that for cases HPC1PPS-SA-N12C_T, PRESSURE SWITCH where specific conditional probabilities have not E41-NO12C SPURIOUSLY ACTUATES been developed, failure or spurious operation is RCI1TME-HI-NO21B_T, TEMPERATURE ELEMENT given a probability of 1.0. E51-TE-N021B SPURIOUS OPERATION RESOLUTION OF CAPABILITY CATEGORY RCI1TME-HI-NO22BT, TEMPERATURE ELEMENT CLASSIFICATION: E51-TE-N022B SPURIOUS OPERATION With the incorporation of the described limited RCI1PPS-SA-NO12A_T, PRESSURE SWITCH changes, BSEP considers the risk results from E51-NO12A SPURIOUS OPERATION the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently RCI1PPS-SA-NO12C_T, PRESSURE SWITCH resolved for SR CF-Al to be assessed as E51-NO12C SPURIOUS OPERATION CAT Il/111 is MET. HPC1PPS-SA-N12B_T, PRESSURE SWITCH E41-NO12B SPURIOUSLY ACTUATES HPC1PPS-SA-N12D_T, PRESSURE SWITCH E41-NO12D SPURIOUSLY ACTUATES SRV1SRV-CO-F013G_T, NON-ADS SAFETY RELIEF VALVE B21-FO13G SPURIOUSLY OPENS RHR1MDP-SA-C002C_T, RHR PUMP E11-CO02C I I BSEP LAR Rev 2 Page V-27

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition SPURIOUS START DUE TO FIRE RCI1PPS-SA-NO12BT, PRESSURE SWITCH E51-NO12B SPURIOUS OPERATION RCI1PPS-SA-NO12D_T, PRESSURE SWITCH E51-NO12D SPURIOUS OPERATION HPC1PPS-SA-N17A_T, PRESSURE SWITCH E41-NO17A SPURIOUS OPERATION HPC1PPS-SA-N17BT, PRESSURE SWITCH E41-NO17B SPURIOUS OPERATION SWS1PPS-SAP129L_T, PRESSURE SWITCH PS129 SPURIOUS OPERATION FAILS LOW ISOLATES HEADER Note that if the instrument spurious operations above are not caused by a hot short, detailed circuit analysis is likely not needed. However, the valve and pump spurious operation would likely benefit from additional analysis. (F&O 2-22) I Page V-28 I I BSEP BSEPLARRev2 LAR Rev 2 Page V-28

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition SF-A2 Seismic Dispositioned BNP-PSA-080, Attachment 17, Section 5 documents The Seismic-Fire Interaction Analysis report has (CAT 1/11/111) Ruggedness and the failure or spurious operation of detection and been updated to address this F&O. SF-A3 Common Cause suppression systems. Flooding, habitability and life Vulnerabilities safety concerns are also addressed, but only through A list of suppression systems that were modified (CAT 1/11/111) reference to the IPEEE. However, no update to this from dry-pipe to wet-pipe systems was determined evaluation is provided. During the walkdown, it was from DBD-62. Using ESR 94-00345, it was noted that some changes in the fire suppression confirmed that the previous flooding analysis system had recently occurred, including changing conducted for the plant remained valid for these suppression systems. Therefore, the modification of some systems from dry to wet-pipe systems. these systems did not introduce any new flooding The following have also not been specifically concerns, and the conclusions from the IPEEE addressed: evaluation remain valid. Discuss seismic vulnerability of any common fire The piping between the diesel and motor driven fire pump suction piping. A common suction for both the pumps is not seismically qualified. Based on electric and diesel fire pump is provided from the drawing review and relevant site documents, a 300,000 gallon storage tank. Failure of this line can single break in the suction piping from the Fire result in failure of both fire pumps. Protection Water Tank or the DWT would not result in the loss of both fire pumps due to the presence of (F&O 3-4) isolation valves. If multiple breaks were to occur due to a seismic event, water supply to both fire pumps could be compromised. DWT suction piping is not considered vulnerable as it is routed underground in some areas. SF-A3 Seismically Dispositioned BNP-PSA-080, Attachment 17, Section 5.2 discusses A) The Seismic-Fire Interaction Analysis report has (CAT 1/11/111) Induced Common common cause suppression failures of the fire water been updated to address this F&O. Cause Failure of system. The common cause failure of gaseous Multiple Fire suppression system (C02 and Halon) is not The Unit 1 and 2 HPCI fire compartments each Suppression discussed. contain an automatic CO 2 suppression system. Systems Each system is supplied by two banks of CO 2 No discussion is provided in regards to establishing supply tanks, designated the main and reserve redundant supply of fire water or gaseous agent banks. These supply tanks are located outside supply. the Reactor Building that they serve. Unit 1 HPCI Fire Compartment FC-RB1-2 is served by Plant procedures should specifically address the main and reserve banks in Fire availability of redundant fire water and gaseous agent supply in case of loss of the main supply of fire water Compartment HCB1, and Unit 2 HPCI Fire Compartment FC-RB2-2 is served by the main I Page V-29 I IBSEP 8SEPLARRev2 LAR Rev 2 Page V-29

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition or normal gaseous agent supply. and reserve banks in Fire Compartment HCB2. (F&O 3-6) Each set of main and reserve banks serves only the automatic suppression system for the adjacent Reactor Building. Based on the close proximity of the main and reserve banks for each system, and their location in a non-seismically qualified fire compartment, a seismic event could damage both the main and reserve supply banks and cause the CO 2 system they supply to become inoperative. However, because the supply for FC-RB1-2 and FC-RB2-2 are separated by a large, open distance, there is no common cause failure that could result in the loss of supply for both automatic CO 2 suppression systems. The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so common cause failure is not a concern. B) The Seismic-Fire Interaction Analysis report has been updated to address this F&O. Discussion of the availability and use of alternate water supply was increased in the report. These alternate supply sources include the DWT and Intake Canal, while the alternate pressure source if both fire pumps are unavailable is an external pump truck. If the fire pumps are unavailable, water supply and pressure can be maintained in the fire suppression ring by external pumper truck through yard hydrants. Each carbon dioxide system for the Unit 1 and 2 HPCI fire compartments contains a main and reserve supply bank, but no other redundant ! I BSEP LAR Rev 2 Page V-30

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition supply was found for these systems. The Unit 1 and 2 systems do not share a common supply and cannot be cross-tied. The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so redundant supply due to common cause failure was not examined. C) The Seismic-Fire Interaction Analysis report has been updated to address this F&O. Plant procedure OOP-41 includes procedures used to align the fire protection system to alternate water supplies and an alternate pressure supply. There is a selector switch for each C02 system to select between main and reserve banks, but no procedure was found for the use of this selector switch. The operation of this selector switch should be included in a procedure to allow for transfer from the main to reserve bank (or vice-versa) in the event the selected supply bank becomes unavailable. No physical cross-tie or procedure to align one unit's HPCI C02 system to the C02 supply of the other unit was found in this analysis. The only compartment under consideration that is equipped with Halon suppression is the Diesel Generator Building Basement (FC-DG-01). The Halon suppressant supply for the system in this fire compartment is local, and so procedures to align a redundant supply due to common cause failure were not examined. I I BSEP LAR Rev 2 Page V-31

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition PP-B1 Justification for Dispositioned Justification for partitioning elements that either lack a Additional justification/clarification was added to (CAT I/Il/Ill) Partitioning fire resistance rating or have been omitted need to be BNP-PSA-083 for the partitioning elements that lack Elements Lacking provided for the following fire compartments (i.e., a fire rating, especially with regard to the presence PP-B2 a Fire Rating examples only): of intervening combustibles for open partitioning (NOT MET) elements. oFC207 - The east wall has an open doorway to PP-C3 FC206 which is not justified RESOLUTION OF CAPABILITY CATEGORY (CAT I/1l/lll) CLASSIFICATION: anFC21 O/FC211 -Tfire rated seals that cannot be maintained as fire barriers With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA

• _FC238 (DG-2) - This compartment also interfaces to be creditable for the NFPA 805 application and with FC244, 245. No justification of partitioning.

this finding to be sufficiently resolved for SR

                                                 -- Generic - Block walls are rated for 2 hours per            PP-B2 to be assessed as CAT Il/111 is MET.

3.2.2.2; however, the walls column identifies them as 3 hours in most cases. Some cases no rating is provided.

                                                  *_FC252   - No justification for unrated block wall -

south.

                                                 *_FC269, 270, 271, 272 - No justification for open grating and stairwell. The only discussion is that openings are beneficial in preventing HGL. If partitioning is not an issue, then it could be combined as one area. Transients or fixed combustible ignition sources and intervening combustibles close to the opening may result in damaging plume temperatures beyond the compartment and/or affect OMAs and fire response.
                                                 *_FC274, 275 - compartment above separated by concrete ceiling and open chase. No justifications for open pipe chase, except that it aids in preventing HGL.
                                                 *_FC278, 279, 284, 285 - Open stairwell, electrical chase and pipe chase are not justified.
                                                 *_FC270 is spatially separated from FC269 by the I

I BSEP LAR Rev 2 Page V-32

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition mezzanine space above HPCI room. No justification has been provided, e.g. distance, intervening combustibles, combustible free zones, etc. (F&O 3-8) FSS-Cl Use of Dispositioned Severity factor calculations are based on generic data Severity factors were applied to every scenario, (CAT I) Conservative per ignition source and the distance to the nearest based on the approved calculation FSS-C4 Severity Factors target (i.e., BNP-PSA-080). Review of the BPNFPRA (NED-M/MECH-1006). However, most had a (CAT I) database (and associated BNPEVAL spreadsheet) Severity Factor of 1.0 because the closest target shows that the distance from the ignition source to the was within the ZOI of the lowest HRR Bin. These nearest target is 0 inches for 3779 of the 4907 distances are based on well documented walk-down sources (including transients). Other target distances results (such that we had a best practice identified are mostly few inches from the source. Resulting SF (F&O 1-33). Sources were typically evaluated for at is 1.0 for almost all scenarios, least two HRRs based on the 75% and the 98% (F&O 3-12) percentile fires. This process has been previously peer reviewed and found to be acceptable. While there may be some conservatism, this is preferable at this point in the process. More detailed analysis is only performed when there is confidence that different results can be obtained which can significantly impact the risk insights. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for both SR FSS-C1 and SR FSS-C4 to be assessed as CAT II is MET. I Page V-33 I I BSEP BSEPLARRev2 LAR Rev 2 Page V-33

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition CS-Al 0 Identification of Dispositioned BSEP cable routing information is contained in the All of the cables routed for FSSPMD contain the (CAT II) Cable Terminal BSEP FSSPMD database. This data base contains terminal information. In fact, the cable naming CS-C2 Data in FSSPMD cable routing information for the selected cables and includes the termination information. There are includes routing information for the analysis unit and some instances where this data has not been (CAT I/Il/Ill) raceway information for the subject cables. The repeated in the FROM/TO fields; however, this field database includes treatment of cable terminal end not required for BSEP. These fields exist because locations for most cables contained in the database. the termination information must be entered However, several cables were found with no terminal specifically in other plants. data included (F&O 4-5) SC-B1 Use of Dispositioned No new thermal hydraulic analysis was used in the New engineering calculations, thermal hydraulic (NOT MET) Engineering construction of the fire PRA; however, there are analysis, and simulator runs were performed to PRM-B7 Judgment Rather several instances where engineering judgment was confirm the success criteria previously established (CATRM Than New used to justify no changes are required in the existing by engineering judgment. There was no change to (/Il/Ill) Thermal success criteria. success criteria previously modeled that were based Hydraulic There are several instances where thermal hydraulic on engineering judgment. The component selection Analysis analysis could analyis have been used to replace eginering engineering calculation to reference(i.e., the BNP-PSA-085) has been specific calculations used updated in judgment in the justification or no justification was determnin the success critia. found for use of existing success criteria in the determining the success criteria. internal events criteria: 1) no evaluation of the affects RESOLUTION OF CAPABILITY CATEGORY on the thermal hydraulic calculation and or timing was CLASSIFICATION: found throughforthe MSO SDVC11-2e (i.e.,drain) vent and RPVfor coolant loss ofdrain 138,000 With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA gal of suppression pool inventory on accident to be creditable for the NFPA 805 application and progression. 2) T23-4U (i.e., Spurious opening of this finding to be sufficiently resolved for SR torus vent and purge valves) no thermal hydraulic SC-B1 to be assessed as CAT II is MET. evaluation of long term affects of short term containment failure on long term containment over pressure. C71-1A (i.e., ATWS) - The justification states that hot shorts may last for up to 11.3 minutes, this may have a significant impact on the thermal hydraulic analysis, and this needs to be considered if this timing is used in the justification for exclusion of the MSO. I I BSEP LAR Rev 2 Page V-34

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition (F&O 4-8) SC-B2 Use of Dispositioned In general analytical methods were not used in the F&O 4-9 seems to confuse expert judgment with (CAT I) Engineering limited changes for success criteria for the fire PRA. conservative decision making, further confounding PRM-B7 Judgment Rather All of the analysis reviewed has some type of the issue by labeling the later as engineering Than Analytical engineering judgment included in the justification. judgment and suggesting that CAT II prohibits its (CAT 1/11/111) Methods For use. Expert judgment is defined in the standard as Changes to MSO P41-5e is an example for a change in the "information provided by a technical expert, in the Success Criteria success criteria of a credited system which includes expert's area of expertise, based on opinion, or on engineering judgments and or assumptions for the an interpretation based on reasoning that includes justification. evaluations of theories, models, or experiments." Case # P41-5e

Description:

Spurious operation (i.e., This differs markedly with the example cited, MSO open) of both RHR service water isolation (i.e., P41-5e, in that the MSO involves a limited number crosstie) valves in a loop may result in diversion of of possible outcomes (i.e., either the flow diversion service water flow from the RHR heat exchangers. fails the NSW pump or it does not). Assuming that the NSW pump fails is certainly the more PRA Disposition: 'Each nuclear service water pump conservative decision. Citing hard data (e.g., pump has an 8,000 gpm design capacity. Each RHR SW design capacity or operating flow rate) for the heat exchanger has a design flowrate of 8,000 gpm. expected performance of specific equipment as the The RBCCW system is adjusted for a 7,200 gpm flow basis for making a conservative decision should not rate. The RBCCW system only automatically isolates cause the resultant stated assumption to be treated on a LOCA or LOOP signal. Since LOCA's are not with the same level of scrutiny as the, presumably considered in a fire PRA, it is assumed that one much softer, information based on an opinion nuclear service water pump is needed and aligned to formed from the evaluation of a theory, model, or RBCCW at the time of the fire. It the second NSW experiment. pump automatically starts (including discharge valve opening) on low NSW header pressure, a spurious RESOLUTION OF CAPABILITY CATEGORY opening of one RHR HX path will be mitigated. If two CLASSIFICATION: or more RHR HX paths spuriously open, it will be assumed that both NSW pumps will fail due to run- With no change being made, BSEP considers the out. Otherwise, if the standby NSW pump does not risk results from the Fire PRA to be creditable for start of its discharge valve does not open, only one the NFPA 805 application and this finding to be RHR HX path needs to be spuriously opened to fail sufficiently resolved for SR SC-B2 to be assessed the operating NSW pump. In this case the standby as CAT Il/111 is MET. NSW pump will also be failed (due to the assumed valve or pump failure). The following combinations model this MSO (and I I BSEP LAR Rev 2 Page V-35

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition SO R43-5j).' (F&O 4-9) FSS-A4 Additional Dispositioned The BSEP approach of fire scenario development The Fire Scenario Data calculation BNP-PSA-086 (NOT MET) Targets for Fire was to evaluate all identified fire sources individually. has been updated to address this Finding. Section Growth Scenarios These fire scenarios included the specific cable tray, 9.5.2 has been updated to include fire propagation. FSS-D1 1 component, and conduit targets for each credible (CAT I/Il/Ill) The database was updated by adding several source. However, review of the information FSS-G1 determined that the identified targets included were queries that create tables which determine the (CAT 1/11/111) only those within the zone of influences of the initial secondary initiator within the most limiting ZOI. All source. No additional targets were included that were other targets that are located above the secondary in the zone of influence for fire growth scenarios initiator (larger DISTV value) are then included to intervening combustibles, such as cable trays were in be in the same ZOI as the limiting secondary the original zone of influence. initiator. This is done by setting the fields 69 kW, 143 kW, 211 kW, 317 kW, and 702 kW for all targets (F&O 4-11) vertically above the limiting secondary initiator to match the same fields of the limiting secondary initiator in the table [Z Source-Target]. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-A4 to be assessed as CAT I/Il/Ill is MET. I Page V-36 I IBSEP BSEPLARRev2 LAR Rev 2 Page V-36

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-D3 Use of Overly Dispositioned The assessment used to quantify the fire risk for the While the assessment is conservative, the Finding (CAT I) Conservative Fire unscreen analysis compartment used fairly contains several factual errors: Modeling Tools conservative approaches, such as consideration of

1) Manual suppression was only applied at 15 only the 75% and 98% fires, inclusion of suppression minutes: Manual suppression timing was based on damage to the first target and/or manual included for all scenarios based on the time to suppression only for time to damage of first target of 15 minutes. damage of the nearest target, and the time to hot gas layer. This is described in the quantification No specific fire modeling, calculations, or analysis calculation and applied in the evaluation were done of the significant fire units analyzed in the spreadsheet.

quantification tasks. More analysis was included for

2) No specific fire modeling analysis was done for the MCR with respect to abandonment; however, the significant fire units: Fire modeling was there still significant conservatism remaining in the calculations such as the below noted in applied to many significant sources. But, as the BNP-PSA-080 'The sensitivity analysis presented in risk of one source decreased, other sources began to dominate the risk.

Appendix B indicates that the fire growth rate and the burning regime can influence the predicted MCR 3) Individual cabinet assessments of fire abandonment times given a peak heat release rate. development and ventilation are needed to However, to fully address these parameters in greater properly assess the time to abandonment detail would require an analysis of individual cabinet conditions: Although the variability of ignition enclosures and an assessment of the fire sites and cable loading/distribution do complicate development and ventilation conditions for each realistic predictions of fire development within cabinet considered.' Therefore only Capability cabinets, the methods used to assess control Category I is considered met. room habitability timing are considered state-of-the-art. (F&O 4-13)

4) CAT II can be MET if the fire risk is bounded.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-D3 to be assessed as CAT II is MET. FSS-E3 Statistical Open BNP-PSA-086, Section 10, contains the identified Additional statistical analysis of the applied heat Representations sources of uncertainty in the fire modeling scope. This release rates and associated parameters is not I BSEP LAR Rev 2 Page V-37

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition (CAT I) of Identified evaluation was limited to a qualitative evaluation of considered practical. Additional statistical analysis Uncertainties the identified uncertainties. No statistical would demonstrate that the conclusions are FSS-H5 representations of the uncertainty intervals was conservative, since bounding inputs provided by (CAT I) present; therefore, only Capability Category I was NUREG/CR-6850 were generally used. The majority FSS-H9 considered met. of applied values are based on the 98th and 7 5 th (CAT 1/11/111) percentile fires from NUREG/CR-6850, and the ZOls The heat release rate, the shortest distance from the are applied conservatively. It is not believed that UNC-A2 ignition source to the target and the fire diameter are reducing these values would allow the use of (CAT 1/11/111) typically considered for statistical representation of reduced impacts for the applications being pursued. uncertainty intervals. The remaining inputs of compartment geometry and ventilation characteristics RESOLUTION OF CAPABILITY CATEGORY are obtained from plant drawings and are typically not CLASSIFICATION: subject to statistical uncertainty analysis. Although no change has yet been made that (F&O 4-14) would improve the Capability Category assessments, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics. FSS-G6 Quantification of Dispositioned BPN-PSA-080 calculation, Section 6, evaluates the Following the methodology in NUREG/CR-6850, (CAT I) Risk for impacts of the MCA evaluations. In Section 6, only BNP-PSA-080 calculation has been revised and the Multicompartment two MCA scenarios were not screened, and required Multi-Compartment Analysis does not assume a Scenarios evaluation. For these two zones, the CCDPs were CCDP of 1.0 for any compartment in the MCA assumed to be 1 and CLERP was assumed to be .1; analysis. Compartment CCDPs were calculated therefore, no specific quantitative evaluations were based on actual localized target sets for exposing performed for these MCAs. As a result Capability compartments. Category Il/111 is not met. RESOLUTION OF CAPABILITY CATEGORY (F&O 4-16) CLASSIFICATION: With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-G6 to be assessed as CAT I/Il/Ill is MET. ! I BSEP LAR Rev 2 Page V-38

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-El Systematic Dispositioned Existing active components identified in the internal A systematic review of the Level 2 progression for (NOT MET) Review of Internal events models were considered in component fire impacts has been performed to the extent Events Accident selection and cable routing. Quantification was required by the design of the core damage FQ-D1 Progression for performed using the existing accident progression sequence models and Level 2 model. This review (CAT 1/11/111) Fire Impacts with no noted changes as related to the affect of fire has been documented as Attachment 14 in the scenarios. Existing modeled operator responses were component selection calculation, BNP-PSA-085, evaluated for changes due to fire affects. The MSO Revision 2. evaluation considered affects of LERF with respect to failure of containment isolation. However, no RESOLUTION OF CAPABILITY CATEGORY systematic review of the accident progression to CLASSIFICATION: determine if fire affects would impact the existing With the incorporation of the described changes, internal events accident progression was found. BSEP considers the risk results from the Fire PRA (F&O 4-17) to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR LE-El to be assessed as CAT 1/11/111 is MET. QU-E3 Accounting for Open Parametric uncertainties that are associated HLR-DA, The state of knowledge correlation has limited (CAT I) the State-of- HR and IE are documented in BNP-0187. However, application to the parameters suggested by this Knowledge the state of knowledge correlation was not considered Finding. A parameter (e.g., fire frequency or non-QU-A3 Correlation in the evaluation of these uncertainty evaluations. suppression probability) in a cutset with no other (CAT II) Correlation should be considered for fire events such similar parameter cannot be correlated. And, the UNC-A1 as the fire frequency, applied severity/HRR split correlation of similar parameters with large failure (NOT MET) fractions, non-suppression, circuit failure probabilities, probabilities (e.g., 0.3 or 0.6 for typical circuit etc. failures) usually yields only fractional increases in FQ-A4 (CAT 1/11/111) risk. Compared to other sources of uncertainty, the (F&O 4-18) correlation of the remaining relevant parameters is expected to yield few additional risk insights. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: Although no change has yet been made that would improve the capability category assessments, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because estimating the uncertainty intervals will not change the quantified risk metrics. I I BSEP LAR Rev 2 Page V-39

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition PP-Al Exclusion of Dispositioned Section 3.2.1 and Attachment 1 of calculation BNP-PSA-083 was revised to add Drywell/Torus, (NOT MET) Areas From the BNP-PSA-083 have been reviewed to examine the Spent Fuel Pool, and VP1NP2 to Global Plant GPAB Based on process by which the Global Plant Analysis Boundary Analysis Boundary. The Spent Fuel Pool and PP-Cl Risk Significance (GPAB) has been defined in the BSEP FPRA. Section (CAT 1/11/111) Service Water Valve Pits were then qualitatively 3.2.1 indicates that all areas that contained any screened, while the Drywell/Torus were simply not PP-C2 equipment or cable credited in the FPRA were quantitatively analyzed based on no fire being (NOT MET) included, as well as any area would require a plant postulated in an inerted atmosphere. shutdown. In addition, any area that is adjacent to an area that would affect FPRA cables/equipment or The characterization of equipment as "risk require a shutdown is said to be included in the significant" was removed from the description of the GPAB. All of these criteria are in agreement with criteria for excluding areas from the GPAB. The PP-Al. distances separating certain buildings of potential interests were added. However, in Attachment 1, a number of buildings/areas are excluded from the GPAB because Distance from ABH to DGB is 32' they do not affect "risk significant" equipment and Distance from CTPH1 to DGB is 28' they may not require a plant shutdown prior to the assumed threshold of 8 hours described in Distance from STORES to RB2 is 30' Assumption 3.1.1.5. This process is consistent with RESOLUTION OF CAPABILITY CATEGORY the guidance provided for the Qualitative Screening CLASSIFICATION: Task (i.e., task 4) in Section 3.3, but is considered inappropriate for use at the PP stage of the analysis With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA All other areas listed in the table in Attachment 1 to be creditable for the NFPA 805 application and should either be confirmed to contain no equipment or this finding to be sufficiently resolved for both cables that are either: SR PP-Al and SR PP-C2 to be assessed as

1) credited in the FPRA (i.e., not just risk significant), CAT 1/11/111 is MET.

or

2) capable of adversely impacting plant response Additionally, the exclusion basis needs to include additional discussion for the following:

                                                 -Aux. Boiler House (ABH) - State that the closest building of concern is the DG building which is approx ft away and will not be affected by an exposure fire in ABH.

oCTPH1 - Due to proximity to DG building, discuss I BSEP LAR Rev 2 Page V-40

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition fire exposure potential.

                                                -Fire house - Any fire alarm panels being affected?
                                                -STORES - Address exposure to south side of the U2 reactor building.
                                                *VP1, VP2 - Not shown on the BGA boundary drawing.

(F&O 5-1) ES-A5 MSO Expert Dispositioned There are a few fire-induced spurious events that Finding 5-4 concerns certain fire-induced spurious (CAT II) Panel Screening were screened, but could in fact either cause a plant events which were screened out by the MSO Expert of Fire-Induced trip (or manual shutdown) and impact equipment that Panel but which, in the opinion of the Fire Peer ES-A6 Spurious Events is credited for accident mitigation in the FPRA: Review Team, could both cause a plant trip or a (CAT II)

1) Spurious start / injection by RCIC. This was Technical Specification mandated manual ES-B2 screened from the FPRA and an initiating event shutdown, and impact equipment that is credited for (CAT II) because it was assumed that no plant trip would accident mitigation in the FPRA.

ES-D1 occur. However, a fire-induced RCIC start would No change will be made to incorporate Finding 5-4 (CAT 1/11/111) likely only be caused if the fire damage was because further consideration of the listed spurious significant enough to cause RCIC inoperability. events revealed no additional fire impacts beyond PRM-B9 Assuming no plant shutdown may be non- what was already identified by the MSO Expert (CAT 1/11/111) conservative. Panel. In particular, contrary to Finding 5-4, the

2) Spurious start / injection by HPCI. This was events described in Items 1, 2, 4, 5, and 6 neither screened from the FPRA and an initiating event cause an automatic plant trip nor require a Technical because it was assumed that no plant trip would Specification mandated manual shutdown in less occur. However, a fire-induced HPCI start would than the 8-hours assumed for treatment as a fire-likely only be caused if the fire damage was induced plant trip.

significant enough to cause HPCI inoperability. Although the event described in Item 3 could either Assuming no plant shutdown may be non- cause an automatic plant trip or prompt the Operator conservative. to initiate a manual scram, depending on the

3) MSO item C11-2e. This MSO drains the RPV number of individual control rods that initially through the SDV vent and drain. The exclusion of scrammed, the RPS scram signal itself would shortly this event from the FPRA is based on the fact that close the SDV vent and drain valves, at which point the suppression pool inventory depletion is slow the scenario would most resemble a previous and would not reach a low enough level in 24 addressed turbine trip.

hours to require a plant trip. However, it may be With regard to the suggested possible resolutions: nonconservative to assume that there is no I Page V-41 I BSEPLARRev2 BSEP LAR Rev 2 Page V-41

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition chance of a plant trip due to this uncontrolled loss For items 1-3, since Section 3.3.2.1 of the MSO of RCS inventory. report (i.e., Attachment 4 of Calculation BNP-PSA-085) already documents significant

4) MSO item E21-001. This MSO describes spurious operator experience for members of the MSO Expert actuation of the core spray pumps and spurious Panel, there is little marginal benefit in citing operation of the injection valves. This event can cause flooding of the main steam lines, which can additional operator interviews for support.

subsequently cause failure of the turbine-driven For item 4, the only clarification necessary would be RCIC/HPCI pumps and EW, which is not to note that the item is incorrectly premised on core modeled. The exclusion justification says that spray being able to inject at high RCS pressure. high-pressure injection is'not credited after depressurization, so there is no way to model the For item 5, the identification in the MSO report of a event. However, if spurious CS pump operation restricting orifice with a 0.105 inch bore should already be sufficient documentation that the HPCI occurred at high RCS pressure and the main steam lines were flooded, HPCI and RCIC should drain pot line to the condenser does not constitute a be impacted because there is still potential for steam flow diversion. crediting their high-pressure injection. For item 6, since an automatic plant trip or manual

5) MSO item E41-2w. This MVSO describes the shutdown is required to drop RPV to below that unisolated drain of HPCI to the main condenser needed for condensate injection, a plot of RPV via spuriously opened AOVs. Two of the three pressure over time is not needed to invalid this MSO AOVs in series have been locked open, so this (i.e., spurious condensate injection with RPV scenario only requires one AOV to open (i.e., on pressure below 500 psig) as an initiating event and loss of instrument air or hot short). This event is would add nothing to the evaluation, in the MSO excluded based on an installed flow-limiting Report (i.e., Attachment 4 of Calculation orifice, but there is no technical discussion of the BNP-PSA-085), of equipment credited for post-trip flow limitation to adequately justify why the accident mitigation.

flowpath is not a valid diversion.

6) MVSO item N21-2ai. This MSO describes RPV overfill due to condensate injection once RPV pressure is <5O0psi. The exclusion justification states that it is unlikely that RCIC/HPCI operation alone would not depressurize the RPV to 500psi in one hour. However, RCIC and HPCI are credited for injection for much longer than 1 hour.

At some point RPV pressure may be reduced to allow condensate injection, which could potentially fail HPCI/RCIC. I I BSEP LAR Rev 2 Page V-42

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition (F&O 5-4) IGN-A4 Strengthen Dispositioned Some of the exclusion bases for the BSEP historical Section 3.4.3 of BNP-PSA-083 was revised to (CAT II) Documentation of fire events should be strengthened to support the include additional discussion of the plant history and Plant Fire History conclusion that the use of generic ignition frequency corrective actions concerning fires related to the IGN-B4 for Use of data is appropriate: heater drain pumps (i.e., Items #1, #2, #5, and #7). (CAT 1/11/111) Generic Ignition The appropriate exclusion of Item #3 as being Frequency Data 1) FR 88-006: A heater drain pump ignited and outside the GPAB was confirmed. The appropriate required three C02 extinguishers at power. exclusion of Item #4 and Item #6 as being not Approximately 2 quarts of oil were burned. This potential challenging was confirmed. Some further appears to be potentially challenging. clarification of the documentation for Items #3, #4,

2) FR 90-002: A heater drain pump ignited at power and #6 was considered, but judged unnecessary at and required 'several' extinguishers. Fire was this time.

fueled by pump oil, caused a fire alarm, and resulted in -$64k worth of damage. This appears to be potentially challenging.

3) FR 94-007: A CWOD pump ignited at power and required offsite fire department response. If this was not dismissed in PP, this could be a potentially challenging fire.

4) ACR 94-01488: A fire in a Rad Waste control room panel at power required a fire extinguisher.

Fire caused loss of SFPC, which appears to be potentially challenging.

5) ACR 97-1136: A heater drain pump ignited at power and was secured to extinguish the fire in response to the fire alarm. This appears to be potentially challenging.

6) ACR 98-651: A cable fire started in a manhole at power due to water intrusion and corrosion. The fire was self-extinguishing, but cable damage was reported.

7) NCR 24699: A heater drain pump ignited an oil fire, which caused a fire alarm and required C02 extinguishers while at power. A condensate I

I BSEP LAR Rev 2 Page V-43

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition system transient resulted and an unusual event was declared due to a duration of >1 0 minutes. This appears potentially challenging. (F&O 5-8) QU-D2 Review for Dispositioned There is no record of a review being performed to A systematic review for modeling inconsistencies (NOT MET) Inconsistencies confirm that the FPRA modeling is consistent from associated with fire impacts was performed. The Between the event sequence to system model or that with approach adopted for including fire-related modeling QU-F3 FPRA and Plant operational characteristics. Since the FPRA model is changes to internal events system logic and (CAT I) Practices That largely based on the internal events model, this is accident sequences was very deliberate at FQ-E1 May Have Been assumed to be a relatively insignificant source of addressing the model capabilities to avoid such (NOT MET) Created by Fire- potential model inaccuracy. However, a review does inconsistencies. The review approach has been Specific Modeling need to be performed to confirm that fire-specific documented in the component selection calculation FQ-F1 Considerations modeling considerations have not created any (i.e., BNP-PSA-085). (NOT MET) inconsistencies between sequence and system modeling, or between the FPRA model and actual RESOLUTION OF CAPABILITY CATEGORY plant operational practices. CLASSIFICATION: With the incorporation of the described changes, (F&O 5-13) BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR QU-D2, SR FQ-E1, and SR FQ-F1 each to be assessed as CAT 1/11/111 is MET and for SR QU-F3 to be assessed as CAT Il/111 is MET. I Page V-44 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-44

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition QU-F3 Review of Non- Dispositioned A review of the cutset review documentation indicates After the peer review, non-significant cutsets were (CAT I) Significant that the vast majority, if not all, of the reviewed reviewed and the results are documented in Cutsets cutsets are from significant scenarios, almost Attachment 39 of BNP-PSA-080, Revision 2 (i.e., QU-D5 exclusively with CCDPs of 1.0. Many of these CCDP Change Package BNP-0235). (NOT MET) cutsets have only a single cutset (i.e.,other applicable FQ-E1 RESOLUTION OF CAPABILITY CATEGORY cutsets are truncated). At the current stage of the (NOT MET) BSEP FPRA development, this is not an CLASSIFICATION: unreasonable characteristic of the cutset reviews. With the incorporation of the described changes, FQ-F1 However, a lack of review of non-significant cutsets BSEP considers the risk results from the Fire PRA (NOT MET) precludes meeting this SR. to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for (F&O 5-14) SR QU-D5, SR FQ-El, and SR FQ-F1 each to be assessed as CAT 1/11/111 is MET and for SR QU-F3 to be assessed as CAT Il/111 is MET. I Page V-45 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-45

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition QU-F2 Identification of Open Although a limited identification of fire CoF Results of the Fire PRA have been reviewed and top (CAT 1/11/111) Significant contributors has been performed, the types of contributors by ignition source, including transient Contributors and contributors is limited and there is little or no and fixed, and compartment have been identified. QU-F3 Review of Risk discussion of the risk insights gained from the Documentation of risk important actions and (CAT I) Importances contributor identification. components has yet to be completed; however, QU-D6 does not affect the outcome of the analysis. Eor example, the following contributions could be (CAT I) insightful, and have not been identified: RESOLUTION OF CAPABILITY CATEGORY QU-D7 CLASSIFICATION: (NOT MET)

                                               -  significant accident sequences Although no change has yet been made that FQ-E1
                                               -  risk significant operator actions performed inside the would improve the Capability Category main control room (NOT MET)                                                                                                 assessments, BSEP considers the risk results FQ-F1                                      -  risk significant operator actions performed outside       from the Fire PRA to be creditable for the (NOT MET)                                    the main control room                                        NFPA 805 application because documenting the significant contributors and importance measures
                                               - contribution to fire COF from transient ignition will not change the quantified risk metrics.

sources

                                               - contribution to fire CDF from fixed ignition sources
                                               - significant spurious actuation events
                                               - significant random failure events (i.e., non-fire),

including common cause failures

                                               - the reduction in ignition frequency contribution to fire CDF due to the extensive use of the conditional plant trip probabilities Additionally, the importance of components and basic events were not reviewed to determine that they make logical sense (QU-D7).

(E&O 5-15) I Page V-46 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-46

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-F1 Identification, Open With respect to identifying the contributors to fire Results of the Fire PRA have been reviewed and top (NOT MET) Review, and LERF, the following contributors are considered: contributors by ignition source and compartment Documentation of have been identified. Documentation of risk LE-F2 - contributions from fire scenarios, MCA Certain LERF important actions and components has yet to be (NOT MET) Contributions, abandonment, and the multi-compartment analysis. completed; however, this does not affect the LE-G3 Importances, and - compartments with >1% fire LERF outcome of the analysis. (CAT I) Uncertainties

                                               -  ignition sources with >1% LERF                       RESOLUTION OF CAPABILITY CATEGORY UNC-Al                                                                                             CLASSIFICATION:

(NOT MET) No identification of plant damage states or containment failure modes was identified, which is Although no change has yet been made that FQ-E1 required for CAT 1.To meet CAT 11,additional would improve the Capability Category (NOT MET) identification of significant fire LERF contributors is assessments, BSEP considers the risk results FQ-F1 required, as discussed in the SR. from the Fire PRA to be creditable for the NFPA 805 application because documenting the (NOT MET) Within the scope of fire LERF contributors that have significant contributors and importance measures been identified, it is not apparent that a review for for LERF will not change the quantified risk

                                                'reasonableness' has been performed.

metrics. For example, 98.1% of Unit 2 fire LERF is due to fires in the Unit 2 main control room. Although this is identified in table 11-2 of BNP-PSA-080, there is no discussion of this considerable contribution including whether or not it is considered reasonable. Notably, the Unit 1 MCR contributes -60% of Unit 1 fire LERF, and no discussion of this asymmetry is provided. (F&O 5-16) I Page V-47 I BSEPLARRev2 BSEP LAR Rev 2 Page V-47

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-G2 Identification of Open Assumptions for the quantification task are Uncertainties in the general Fire PRA model and (NOT MET) Sources of documented in Section 3.3 of BNP-PSA-080. General quantification have been identified in the Fire PRA LE-F3 Uncertainty sources of uncertainty are discussed in Section 8.4. Quantification calculation and in the fire PRA (NOT MET) Peculiar to Fire These sources include: development calculations. These sources of LERF and uncertainty are valid in the fire LERF and fire CDF

                                                - ignition frequencies LE-G4       Documentation of                                                                       quantifications. No additional LERF specific sources LERF Importance               - HRRs                                                   of uncertainty have been identified, and importance (NOT MET)

Measures and - target selection measures and statistical analysis of LERF have not UNC-A1 Statistical - damage time yet been performed. (NOT MET) Uncertainty - time to HGL

                                                - fire effects                                           RESOLUTION OF CAPABILITY CATEGORY UNC-A2
                                                - suppression                                            CLASSIFICATION:

(CAT 1/11/111) - circuit analysis Although no change has yet been made that FQ-E1 - HRA would improve the Capability Category (NOT MET) - quantification (including tools) assessments, BSEP considers the risk results FQ-F1 These sources of uncertainty are valid in the fire from the Fire PRA to be creditable for the (NOT MET) LERF and fire COF quantifications, but there are no NFPA 805 application because documenting additional sources of uncertainty that are applicable to additional sources of uncertainty and importance the fire LERF calculation. Change package BNP-01 87 measures for LERF will not change the quantified provides fire COF importance measures and a risk metrics. statistical analysis of fire COF uncertainty, but does not address fire LERF. (F&O 5-18) I Page V-48 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-48

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-G6 Definition of Dispositioned There is no definition established for 'significance' A discussion of "significance" in terms of the (NOT MET) "Significance" related to basic events, cutsets, accident sequences, definitions described in ASME/ANS-Ra-Sa-2009 or any other facets of the fire PRA results. Section 1-2 has been added to Section 8 of the QU-F6 Quantification Calculation (i.e., BNP-PSA-080). (NOT MET) (F&O 5-19) FQ-F1 RESOLUTION OF CAPABILITY CATEGORY (NOT MET) CLASSIFICATION: With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR LE-G6, SR QU-F6, and SR FQ-F1 each to be assessed as CAT 1/11/111 is MET. I Page \~-49 IBSEPLARRev2 I BSEP LAR Rev 2 Page V-49

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition LE-G2 Documentation of Dispositioned The deficient sub-requirements of this SR are detailed Attachment 13 was added to BNP-PSA-085, (NOT MET) the LERF below. Revision 2, to address items A, B, D, E, F, and part Analysis (i.e., equipment, containment failure modes and FQ-F1 A) No documentation was provided of plant damage phenomena) of C. (NOT MET) states / attributes, although this can be considered covered by general references to the The remainder of item C (i.e., fire-specific human internal events PRA model. actions considered in the fire LERF sequence development) was addressed in Section 4.2.3 and B) There is no documentation of how accident Table 5.1 of BNP-PSA-084, Revision 2. sequences were binned into plant damage states, but since the fire LERF model is based on the F&O 5-18 sufficiently addresses item H (i.e., internal events LERF model, references to the LERF-related uncertainty). Resolution will be internal events PRA can account for this. completed as part of F&O 5-18. C) There should be discussion of the fire-specific RESOLUTION OF CAPABILITY CATEGORY human actions and equipment considered in the CLASSIFICATION: fire LERE sequence development. Containment With the incorporation of the described changes, failure modes and phenomena could be BSEP considers the risk results from the Fire PRA referenced to the internal events documentation to be creditable for the NFPA 805 application and D) There is no discussion of fire-specific factors this finding to be sufficiently resolved for both influencing containment challenges and SR LE-G2 and SR FQ-F1 to be assessed as containment capability. CAT 1/11/111 is MET. E) Containment capacity analysis could be covered by a reference to the internal events LERF model. No fire-specific impacts are expected. F) A discussion of fire-specific impacts on the accident sequences identified in the containment event trees should be provided. H) The model integration process is described in Section 4.9 of BNP-PSA-080. There is no fire LERF-related uncertainty (F&O 5-18) or sensitivity analyses provided. (F&O 5-20) ! Page V-50 I IBSEPLARRev2 BSEP LAR Rev 2 Page V-50

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition CS-B1 Review of Dispositioned There is evidence that the existing electrical Evidence of the collective review of electrical (CAT I) Electrical coordination analysis was reviewed and refined (i.e., coordination with supporting analysis for breakers, Coordination for BNP-0157). Specific documentation should be power supplies, and cables was documented in CS-C4 provided of this review. There is no evidence that Credited Power Revision 2 of BNP-PSA-080 as: (NOT MET) Supplies power supplies credited in the fire PRA were reviewed Attachment 13 (i.e., Change Package BNP-0157) to confirm that they were addressed by existing overcurrent calculations. Attachment 36 (i.e., Change Package BNP-0218) (F&O 6-1) Attachment 37 (i.e., Change Package BNP-0224) Attachment 41 (i.e., Change Package BNP-0215) Attachment 42 (i.e., Change Package BNP-0217) Attachment 43 (i.e., Change Package BNP-0223) Attachment 44 (i.e., Change Package BNP-0224). RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With the incorporation of the described changes, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR CS-B1 to be assessed as CAT Il/111 is MET and for SR CS-C4 to be assessed as CAT I/llI/ll is MET. I Page V-51 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-51

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-G4 Failure Probability Dispositioned Passive fire barriers with a fire resistance rating are Per the industry training and practice, it was never (CAT I) for Barrier With credited in the multicompartment analysis. The failure meant for the probabilities for individual elements Multiple Rated rates used are those prescribed in NUREG-6850, over an entire wall to be summed to get probability Elements however, the worst case value for failure probability of of wall failure. Summing the probabilities implies the the barrier is used. failures of individual barriers have no dependence and, in walls with multiple penetrations and barriers, (F&O 6-4) could result in a barrier failure probability greater than 1. Walkdowns were performed to gather the targets and barriers between the exposing and exposed compartments. The worst case barrier failure probability was applied to all local targets between two adjacent compartments. The results of this analysis are included in Rev 1 of the BSEP fire quantification calc. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-G4 to be assessed as CAT 11is MET. FSS-G2 Consideration of Dispositioned Screening methodology is provided in BNP-PSA-080, Plant walkdowns were performed to identify targets (CAT 1/11/111) Localized Effects Section 6.0. in the exposed compartments near the barriers For separating the exposing and exposed Multicompartment However,: the MCA screening did not consider the impact of possible localized effect (i.e., damage to compartments. The fire quantification calculation Screening Criteria was revised to include the localized damage in the equipment) near penetrations and barriers. adjacent compartment near barriers for all In addition, a screening value was used without compartments that screened out and for justification and the cumulative risk for the screened compartments where MCA was performed but did scenarios was not evaluated. not achieve a HGL in the combined compartments. The localized targets of the adjacent compartment (F&O 6-5) were added to the HGL evaluation for the exposed compartment. I Page V-52 I IBSEP LAR Rev 2 BSEPLARRev2 Page V-52

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition CF-Al Application of Dispositioned Conditional failure probabilities were assigned to No non-conservative application of conditional (CAT I) Conditional selected cables per the methodology identified in failure probabilities has been identified for an off-Failure BNP-01 37, which is based on the Chapter 10 tables scheme cable. Probabilities to in NUREG-6850. However the BSEP methodology for For safe shutdown, any failure of an associated Off-scheme determining the component level spurious operation Cables probability, as identified in BNP-PSA-080 Section circuit also fails the main component. This is 4.3.4 and 4.6.1.2.4, is to use the worst case spurious conservative, in that typically only one or two of an operation probability of all affected cables without associated circuit's cables actually affect the primary regard as to whether the cables in question are component. When applied to the Fire PRA, this primary scheme or off-scheme cables. Per method of including associated circuits created far FAQ 08-047, off-scheme cables and cables with too many false failures and, therefore, associated alternate source breakers must be identified and, circuits are not always linked to the primary when combined with on-scheme cables, an exclusive component as shown in FSSPMD. In almost all OR must be used. Spurious events of high cases, the associated circuits are modeled importance that had spurious operabilities applied separately as primary components in the fire PRA were reviewed and found to have no off-scheme fault tree. In this manner, cable damage to the cables; therefore, CAT I is considered met. associated circuit is captured within the fault tree, and will cause cascading failures based on the (F&O 6-7) model. In addition, key interlocks that can have an impact on the Fire PRA are included in the model. Therefore, it can be determined that off-scheme cables are included. Additionally, many times, although they are included, the failure probability may be 1.0, and appear to be unanalyzed. In assigning the fault probabilities for Brunswick, specific basic events were identified by PRA. Fault probabilities were assigned to the on-scheme cables that could affect the basic event of concern. The values assigned represented the best-estimate as shown in the tables in Chapter 10 of NUREG/CR-6850. These fault probabilities were in general, only applied to control circuits. A loss of power that results in the failure of a basic event could occur due to a short to ground, and since the fault probabilities provided in NUREG/CR-6850 only apply to hot shorts, a probability of 1.0 would be assigned. Similarly, I I BSEP LAR Rev 2 Page V-53

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition instrumentation cables are assumed to fail with a probability of 1.0 since they have not been specifically tested. However, since they can fail either high or low, a split fraction may still be applied to the functional response to the cable fault. Since many of the associated circuits are tied to instrumentation, not performing a fault probability analysis on such circuits has no impact on the PRA results since the failure would be an assumed value of 1.0, and no advantage would be gained. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR CF-Al to be assessed as CAT Il/111 is MET. ! Page V-54 I I BSEP LAR Rev 2 BSEPLARRev2 Page V-54

CP&L Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II) SR Topic Status Finding Disposition FSS-A1 Split Fraction for Dispositioned (Unreviewed Analysis Method) BSEP does not consider this to be an unreviewed (NOT MET) "Open"/"Closed" The BSEP FPRA calculates using: analysis method because this treatment is described MCCs in a sensitivity study in Section 3.4.7 and

1) A severity factor 0.1, where 90% of the fires are Table 3.4-6 of the Safety Evaluation for the HNP contained within the MCC license amendment (ML101130535). In particular, the assumption that a small percentage of fires will

2) HRR severity factors are treated independently, cause damage outside the MCC cabinet was similar to other cabinets. identified with an assessment of the physical design (F&O 4-1) and associated fire modeling as a reasonable basis for considering the MCCs as closed cabinets.

Following the guidance provided by NUREG/CR-6850 it has been determined that some MCCs can be treated as "closed" cabinets. As such, there is no impact to external targets. Based on challenges that there is potential for an arc fault to have enough energy to open the cabinet, even though the documentation specifically excludes the need to apply HEAFs to MCCs, it is assumed that one out of ten MCC fires may result in an "open" cabinet configuration. This is not applied to the HRR as a severity factor, but as a split fraction on the likelihood on the cabinet to be "closed". RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION: With no change being made, BSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-A1 to be assessed as "CAT 1/11/111 is MET." I Page V-55 I IBSEP LAR Rev 2 BSEPLARRev2 Page V-55}}