View source for L-2012-092, License Amendment Request No. 216, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition). Part 1 of 4

{{Adams
| number = ML12191A048
| issue date = 06/28/2012
| title = License Amendment Request No. 216, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition). Part 1 of 4
| author name = Kiley M
| author affiliation = Florida Power &amp; Light Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000250, 05000251
| license number = 
| contact person = 
| case reference number = L-2012-092
| document type = Letter type:L, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 498
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:0   FPLO                               Florida Power & Light Company, 9760 SW 344th St., Homestead, FL 33035 June 28, 2012 10 CFR 50.90 POWERING TODAY.
EMPOWERING TOMORROW.                                                                       L-2012-092 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Re:      Turkey Point Nuclear Generating Station Units 3 and 4 Docket Nos. 50-250 and 50-251 License Amendment Request No. 216 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
In accordance with the provisions of 10 CFR 50.90, "Application of License or Construction Permit," Florida Power & Light Company (FPL) hereby requests an amendment to the Renewed Facility Operating License (RFOL) for Turkey Point Nuclear Generating Station Units 3 and 4.
Implementation of the regulatory actions presented in the enclosure to this License Amendment Request (LAR) will enable FPL to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision I of Regulatory Guide (RG) 1.205.
As documented in this request, FPL has met all of the regulatory requirements for the transition of its fire protection licensing basis. This license amendment does not present a significant hazards consideration and the criteria have been met for categorical exclusion from the need for an environmental assessment.
Information associated with the transition of Turkey Point Nuclear Generating Station Units 3 and 4 is provided in the enclosed Transition Report, including the technical and regulatory justifications required to support this LAR. Major topics in this Transition Report are presented as Attachments A through W, summarized as follows:
Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.
Attachments C, D, G, S and W of the Enclosure to this letter contain security-related information. Upon removal of Attachments C, D, G, S and W of the enclosure, this letter is uncontrolled.
an FPL Group company

Florida Power & Light Company                                                       L-2012-092 Turkey Point Nuclear Generating Station Units 3 and 4                               Page 2 of 4 License Amendment Request 216 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
* Attachment A - NEI 04-02 Table B Transition of Fundamental FP Program and Design Elements (NFPA 805 Chapter 3)
    "   Attachment B - NEI 04-02 Table B Nuclear Safety Capability Assessment -
Methodology Review
    "   Attachment C - NEI 04-02 Table B Fire Area Transition - Security Related Information
    "   Attachment D - NEI 04-02 Table F-I Non-Power Operational Modes Transition -
Security Related Information
* Attachment E - NEI 04-02 Table G- I - Radioactive Release Transition
    "   Attachment F - Fire-Induced Multiple Spurious Operations Resolution
    "   Attachment G - Operator Manual Actions - Security Related Information
* Attachment H - NFPA-805 Frequently Asked Question Summary Table
    "   Attachment I - Definition of Power Block
* Attachment J - Fire Modeling V & V
    "   Attachment K - Existing Licensing Action Transition
    "   Attachment L - NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
* Attachment M - License Condition Changes
    "   Attachment N - Technical Specification Changes
* Attachment 0 - Orders and Exemptions
    "   Attachment P - Risk-Informed, Performance-Based Alternatives to NFPA 805
    "   Attachment Q - No Significant Hazards Evaluation
    "   Attachment R - Environmental Considerations Evaluation
* Attachment S - Plant Modifications and Items to be Completed During Implementation -
Security Related Information
    "   Attachment T - Clarification of Prior NRC Approvals
    "   Attachment U - Internal Events PRA Quality
* Attachment V - Fire PRA Quality
    "   Attachment W - Fire PRA Insights - Security Related Information Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.
Attachments C, D, G, S and W of the Enclosure to this letter contain security-related information. Upon removal of Attachments C, D, G, S and W of the enclosure, this letter is uncontrolled.

Florida Power & Light Company                                                         L-2012-092 Turkey Point Nuclear Generating Station                                               Page 3 of 4 License Amendment Request 216 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
In accordance with the guidance in Regulatory Position 2.2.4.2 of Regulatory Guide 1.205, Revision 1, the total risk change associated with pre-transition fire protection program variances meeting the NFPA 805 performance-based approach (via the risk evaluation process) was evaluated. Upon completion of the plant modifications, as referenced in the Transition Report Section 4.8.2, the total change in risk associated with the transition of the Turkey Point Nuclear Generating Station Units 3 and 4 to NFPA 805 will be consistent with the acceptance guidelines in Regulatory Guide 1.174.
The Transition Report Section 5.5 contains FPL's proposed implementation schedule for transitioning to the new fire protection licensing basis. The Licensee Commitments, identified in Tables S-2 and S-3 of Attachment S of the enclosed Transition Report, are proposed to support approval of the LAR and may change during the NRC review period.
FPL requests that Attachments C, D, G, S and W to the enclosure, which contain sensitive security-related information, be withheld from public disclosure in accordance with 10 CFR 2.390.
The Plant Nuclear Safety Committee has reviewed the proposed amendment. In accordance with 10 CFR 50.91, a copy of this application, with attachments and enclosure, is being provided to the designated official of the State of Florida.
The following schedule for transitioning Turkey Point Nuclear Generating Station Units 3 and 4 to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule:
     " Implementation of the 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 unless that falls within a scheduled outage window. Then this will occur 60 days after startup from that scheduled outage. See Attachment S Table S-3.
     " Modifications will be completed by the startup of the second refueling outage (for each unit) after the issuance of the SER. Appropriate compensatory measures will be maintained until modifications are complete. See Attachment S, Table S-2.
Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.
Attachments C, D, G, S and W of the Enclosure to this letter contain security-related information. Upon removal of Attachments C, D, G, S and W of the enclosure, this letter is uncontrolled.

Florida Power & Light Company                                                       L-2012-092 Turkey Point Nuclear Generating Station                                             Page 4 of 4 License Amendment Request 216 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
If you should have any questions regarding this application, please contact Robert Tomonto, Licensing Manager, at 305-246-7327.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on June 28, 2012.
Very truly yours, Michael Kiley Vice President Turkey Point Nuclear Generating Station

==Enclosure:==
FPL's Turkey Point Units 3 and 4 Transition Report cc:     Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, Turkey Point USNRC Project Manager for Turkey Point Mr. W. A. Passetti, Florida Department of Health Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390.
Attachments C, D, G, S and W of the Enclosure to this letter contain security-related information. Upon removal of Attachments C, D, G, S and W of the enclosure, this letter is uncontrolled.

Enclosure Florida Power & Light Company Turkey Point Nuclear Generating Station Units 3 and 4 License Amendment Request No. 216 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

Florida Power & Light Company Turkey Point Nuclear Generating Station Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition FPLo Transition Report June 29, 2012

Florida Power & Light                                                                               NFPA 805 Transition Report TABLE OF CONTENTS

==1.0    INTRODUCTION==
 .....................................................................................................             1 1.1 Ba ckg ro und ..................................................................................................          .. 1 1.1.1 NFPA 805- Requirements and Guidance ............................................                                       1 1.1.2 Transition to 10 CFR 50.48(c) ...............................................................                          2 1.2 P urp o se .............................................................................................................      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 ...................................................................................                            7 3 .1 Ba ckg ro und ..................................................................................................         .. 7 3.2 NFPA 805 P rocess ......................................................................................                  .. 7 3.3 NEI 04 NFPA 805 Transition Process ...................................................                                   8 3.4 NFPA 805 Frequently Asked Questions (FAQs) ............................................                                       9 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS ........................................                                                   11 4.1 Fundamental Fire Protection Program and Design Elements ...................... 11 4.1.1 Overview of Evaluation Process ..........................................................                              11 4.1.2 Results of the Evaluation Process .......................................................                              13 4.1.3 Definition of Power Block and Plant .....................................................                              14 4.2 Nuclear Safety Performance Criteria ............................................................                             14 4.2.1 Nuclear Safety Capability Assessment Methodology ...........................                                           14 4.2.2 Existing Engineering Equivalency Evaluation Transition ..................... 21 4.2.3 Licensing Action Transition .................................................................                          22 4.2.4 Fire Area Transition ............................................................................                      26 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 ..................................................                                  31 4.4.1 Overview of Evaluation Process ..........................................................                              31 4.4.2 Results of the Evaluation Process .......................................................                              32 4.5 Fire PRA and Performance-Based Approaches .........................................                                          32 4.5.1 Fire PRA Development and Assessment ............................................                                       32 4.5.2 Performance-Based Approaches ........................................................                                  34 4 .6 Monitoring P rogram ...................................................................................                 . . 38 Page Revision 0 Revision   0                                                                                                                     Page i

Florida Power & Liqht                                                                                  NFPA 805 Transition Report 4.6.1  Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program ...... 38 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ............... 39 4.7   Program Documentation, Configuration Control, and Quality Assurance ........ 44 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 8 0 5 ....................................................................................................       . . 44 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2 .9 of NFPA 805 ..............................................................................                  . 47 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 .... 50 4.8   S um m ary of R esults ..................................................................................            . . 52 4.8.1 Results of the Fire Area Review .........................................................                              52 4.8.2 Plant Modifications and Items to be Completed During the Implementation P h a se ................................................................................................       . . 53 4.8.3 Supplemental Information - Other Licensee Specific Issues ............... 54

==5.0   REGULATORY EVALUATION==
 .........................................................................                                55 5.1   Introduction - 10 C FR 50.48 ........................................................................                      55 5.2   R egulatory T opics ......................................................................................             . . 60 5.2.1 License Condition Changes .................................................................                            60 5.2.2 Technical Specifications .....................................................................                         60 5.2.3 Orders and Exemptions .....................................................................                            60 5.3   Regulatory Evaluations ...............................................................................                     60 5.3.1 No Significant Hazards Consideration ................................................                                  60 5.3.2 Environmental Consideration ...............................................................                            61 5.4 Revision to the PTN UFSA R ..........................................................................                        61 5.5 Transition Implementation Schedule ............................................................                              61

==6.0   REFERENCES==
 ..................................................................................................                  62 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 R eview ................................................................................................................      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 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 Revision 0                                                                                                                      Page ii

Florida Power & Light                                                                        NFPA 805 Transition Report 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. Plant Modifications and Items to be Completed During Implementation ...... 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 Page iii Revision 0 Revision  0                                                                                                            Page iii

Florida Power & Light                                                   Executive Summary Executive Summary Florida Power & Light will transition the Turkey Point Nuclear Plant Units 3 and 4 fire protection program to a new Risk-Informed, Performance-Based alternative per 10 CFR 50.48(c) which incorporates by reference NFPA 805. The licensing basis per 10 CFR 50.48(b) and 10 CFR 50, Appendix R will be superseded.
The voluntary adoption of 10 CFR 50.48(c) by Turkey Point Nuclear Plant does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, General Design Criteria 3, Fire Protection. Compliance with the new regulation establishes compliance with these sections.
Florida Power & Light submitted a letter of intent to the Nuclear Regulatory Commission on November 15, 2005 for Turkey Point Nuclear Plant Units 3 and 4 to adopt National Fire Protection Association Standard 805 (NFPA 805) in accordance with 10 CFR 50.48(c). By letter dated September 18, 2006, the NRC granted a three year enforcement discretion period. In accordance with COMSECY-08-022, Florida Power &
Light requested on September 11, 2008 that the due date for the 10 CFR 50.48(c) license amendment request be extended to 6 months past the date of the safety evaluation approving the second pilot plant. The Nuclear Regulatory Commission granted the enforcement discretion extension on November 14, 2008. In accordance with SECY-1 1-0061, Florida Power & Light requested on June 23, 2011 that the due date for the 10 CFR 50.48(c) license amendment request be extended to June 30, 2012. The Nuclear Regulatory Commission granted the enforcement discretion extension on August 8, 2011. The enforcement discretion period will continue until the NRC issues a License Amendment.
This submittal incorporates lessons learned from the Requests for Additional Information for the pilot plants, and the generic NFPA 805 License Amendment Request template developed by the Nuclear Energy Institute NFPA 805 Task Force.
The transition process consisted of a review and update of Turkey Point Nuclear Plant documentation, including the development of a Fire Probabilistic Risk Assessment Model 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 04-02 (NEI 04-02) Revision 2 and applicable Frequently Asked Questions, and
* 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 Revision 0                                                                          Page iv

Florida Power & Light                                                   Executive SurnmarV Radioactive Release Performance Criteria,                                          0
* Monitoring Program, and
* 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, and
* Determination of No Significant Hazards and evaluation of Environmental Considerations.
The attachments to the Transition Report include detail to support the transition process and results.
Attachment H contains the approved Frequently Asked Questions not yet incorporated into the endorsed revision of NEI 04-02. These Frequently Asked Questions have been used to clarify the guidance in Regulatory Guide 1.205, NEI 04-02, and the requirements of NFPA 805.
Revision 00                                                                         Page v Revision                                                                            Page v

Florida Power & LiQht                                               Acronym List Acronym List AC                   alternating current ADAMS                Agencywide Documents Access and Management System ADS                  automatic depressurization system ADV                  atmospheric dump valve AFW                  auxiliary feedwater AHJ                  authority having jurisdiction AHU                  air handling unit AMSAC                ATWS mitigation system actuation circuitry ANS                  American Nuclear Society ANSI                 American National Standards Institute AOP                  abnormal operating procedure AOV                  air operated valve APCSB                Auxiliary Power Conversion Systems Branch ASD                  alternate shutdown ASME                 American Society of Mechanical Engineers ASEP                 Accident Sequence Evaluation Program ASP                  auxiliary shutdown panel ASTM                 American Society of Mechanical Engineers ATWS                 anticipated transient without scram BAT                  boric acid tank BTP                  Branch Technical Position BTU                  british thermal unit BWR                  boiling water reactor BWROG                boiling water reactor owners group CC                   capability category Page vi Revision 00 Revision                                                                  Page vi

Florida Power & Light                                             Acronym List CCDP                 conditional core damage probability CCF                  common cause failure CCW                  component cooling water CDF                  core damage frequency CF                   ceramic fiber CFAST                Consolidated Fire and Smoke Transport CFR                  Code of Federal Regulations CLB                  current licensing basis Cm                   centimeter CR                   control room CRDM                 control rod drive mechanism CS                   control switch CSD                  cold shutdown CT                   current transformer CVCS                 chemical and volume control system DBA                  design basis accident DBD                  design basis document DC                   direct current DDFP                 diesel-driven fire pump DH                   decay heat DID                  defense-in-depth DOE                  Department of Energy ECC                  emergency core cooling EDG                  emergency diesel generator EDS                  electrical distribution system EEEE                 existing engineering equivalency evaluation Page vii Revision 0 Revision 0                                                             Page vii

Florida Power & Light                                             Acronym List Florida Power & Light                                             Acronym List EEL                  essential equipment list EF                   error factor EOP                  emergency operating procedure EPRI                 Electrical Power Research Institute EPU                  extended power uprate ERFBS                electrical raceway fire barrier system ESFAS                engineered safety features actuation signal OF                   degrees Fahrenheit F&O                  fact and observation FA                   fire area FAQ                  frequently asked question FCV                  flow control valve FHA                  fire hazards analysis FIF                  fire ignition frequency FP                   fire protection FPER                 fire protection evaluation report FPL                  Florida Power & Light FP&L                 Florida Power & Light FPRA                 Fire Probabilistic Risk Assessment FOT                  fuel oil transfer FR                   Federal Register FSAR                 Final Safety Analysis Report Ft                   feet FZ                  fire zone GDC                  General Design Criterion GL                   U.S. NRC Generic Letter Page viii Revision 00                                                            Page viii

Florida Power & Light                                                  Acronym List Florida Power & Light                                                  Acronym List GPM                  gallons per minute HBr                  hydrogen bromide HCV                  hand control valve HDPE                 high density polyethylene HEAF                 high energy arcing fault HEP                  human error probability HF                  hydrogen fluoride HGL                 hot gas layer HHSI                 high head safety injection HPME                high pressure melt ejection HRA                 human reliability analysis HRE                 higher risk evolution HRR                 heat release rate HSD                 hot shutdown HSS                  high safety significant HVAC                heating, ventilation, and air conditioning HX                  heat exchanger IA                  instrument air ICW                 intake cooling water IE                  internal events IEEE                 Institute of Electrical and Electronic Engineers IN                   U.S. NRC Information Notice IPE                  Individual Plant Examination IPEEE                Individual Plant Examination of External Events ISFSI                Independent Spent Fuel Storage Installation ISLOCA               inter-system loss of coolant accident Page ix Revision 0 Revision 0                                                                  Page ix

Florida Power & Light                                   Acronym List Flrd oe     ih                                     AcoymLs Kg                   kilogram kJ                   kilojoules KSF                  key safety function KV                   kilovolt L                    liter LA                   licensing action LAR                  license amendment request Lb                   pound LC                   load center LERF                 large early release frequency LOCA                 loss of coolant accident LOOP                 loss of offsite power LP                   low pressure LPCI                 low pressure coolant injection LSS                  low safety significant MAAP                 Modular Accident Analysis Program MCA                  multi-compartment analysis MCC                  motor control center MCR                  main control room MFW                  main feedwater MHIF                 multiple high impedance fault MG                   motor generator Min                  minute MOV                  motor operated valve MS                   main steam MSIV                 main steam isolation valve Page x Revision 00 Revision                                                       Page x

Florida Power & Light                                                Acronym List MSO                  multiple spurious operation N2                   nitrogen NEI                  Nuclear Energy Institute NFPA                 National Fire Protection Association NICAD                NISYS Compliance Assessment Database NIST                 National Institute of Standards and Technology NPO                  non-power operational NRC                  Nuclear Regulatory Commission NRHX                 non-regenerative heat exchanger NSCA                 nuclear safety capability assessment NSEL                 nuclear safety equipment list NUMARC               Nuclear Management and Resources Council NUREG                document prepared by the NRC staff NUREG/CR             document prepared by NRC contractors OE                   operating experience OMA                  operator manual action OOS                  out of service P&ID                 piping and instrumentation diagram PAU                  physical analysis unit PB                   performance-based PCS                  primary control station PORV                 power operated relief valve POS                  plant operational state POV                  power operated valve PRA                  probabilistic risk assessment PSA                  probabilistic safety assessment Page xi Revision 00 Revision                                                                  Page xi

Florida Power & Light                                         Acronvmn List Flrd oe     igtArnmLs PSF                   performance shaping factor PTN                   Turkey Point Nuclear Plant PTS                   pressurized thermal shock PWR                   pressurized water reactor PWROG                 pressurized water reactor owners group PVC                   polyvinyl chloride QA                    quality assurance RAW                   risk achievement worth RCA                   radiologically controlled area RCO                   reactor control operator RCP                   reactor coolant pump RCS                   reactor coolant system RG                    U.S. NRC Regulatory Guide RHR                   residual heat removal RI-PB                 risk-informed, performance-based RIS                   Regulatory Issue Summary RO                    reactor operator RPS                   reactor protection system RWST                  reactor water storage tank SAMG                  severe accident management guideline SDC                   shutdown cooling SER                   safety evaluation report SFP                   spent fuel pool SFPE                  Society of Fire Protection Engineers S/G                   steam generator SG                    steam generator Page xii Revision 00 Revision                                                            Page xii

Florida Power & Liqht                                          Acronym List Flrd oe     iqtArnmLs SGFP                  steam generator feed pump SGTR                  steam generator tube rupture SI                    safety injection SM                    safety margin SP                    special publication SR                    supporting requirement SRV                   safety relief valve SSA                   safe shutdown analysis SSC                   structures systems, and components SSE                   safe shutdown earthquake SSEL                  safe shutdown equipment list SSLD                  safe shutdown logic diagram SSGFP                 standby steam generator feedwater pump SSGFW                 standby steam generator feedwater SWGR                  switchgear T&M                   testing and maintenance T-H                   thermal-hydraulic TS                    test switch UFSAR                 Updated Final Safety Analysis Report UV                    under-voltage VAC                   volts - alternating current VCT                   volume control tank VDC                   volts - direct current VFDR                  variance from deterministic requirement W                     watts WOG                   Westinghouse Owners Group Page xiii Revision 00                                                        Page xiii

Florida Power & Light                    Acronym List Flrd oe     iqtArnmLs XS                    transfer switch Yr                    year ZOA                   zone of analysis ZOI                   zone of influence Page xiv Revision 0 Revision   0                                  Page xiv

Florida Power & Light                                                                  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). Florida Power & Light (FPL) 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) (NEI 04-02), to transition Turkey Point Nuclear Plant (PTN) 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 PTN 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:
1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.
Revision 0                                                                                       Page 1

Florida Power & Light                                                              1.0 Introduction Florida Power & Light                                                              1.0 Introduction NFPA 805                                       10 CFR Incorporation by 2001 ed.            Reference                 50.48(c) 4)
E a)                 Performance-Based Standard for FP for                               National Fire Light Water Reactor                                 Protection Electric Generating                                Association Plants                                    Standard NFPA 805 I
                                                                                            -~
NEI 04-02                                  RG 1.205 Endorsement N          .
GUIDANCE 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 FPL submitted a letter of intent to the NRC on November 15, 2005 (ML053290175) for PTN to adopt NFPA 805 in accordance with 10 CFR 50.48(c).
By letter dated September 18, 2006 (ML062020265), 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.
In accordance with COMSECY-08-0022, FPL requested on September 11, 2008 (ML082600482) that the due date for the 10 CFR 50.48(c) license amendment request be extended to 6 months past the date of the safety evaluation approving the second pilot plant. The NRC granted the enforcement discretion extension on November 14, 2008 (ML083180165).
In accordance with SECY-1 1-0061, FPL requested on June 23, 2011 that the due date for the 10 CFR 50.48(c) LAR be extended to June 30, 2012. The NRC granted the enforcement discretion extension on August 8, 2011 (ML11200A191). The enforcement discretion period will continue until the NRC issues a License Amendment.
Revision 0                                                                                     Page 2

Florida Power & Liciht                                                        1.0 Introduction 1.1.2.2     Transition Process The transition to NFPA 805 includes the following high level activities:
* Revision to the existing fire safe shutdown analysis,
* A new Fire Probabilistic Risk Assessment (PRA), and M 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.
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, and
: 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.
Page 3 Revision 00                                                                             Page 3

Florida Power & Light                          2.0 Overview of Existing Fire Protection Program 2.0     OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM 2.1     Current Fire Protection Licensing Basis PTN Unit 3 was licensed to operate on July 19, 1972 and PTN Unit 4 was licensed to operate on April 10, 1973. As a result, the PTN fire protection program is based on compliance with 10 CFR 50.48(a), 10 CFR 50.48(b), and the following License Condition:
FPL's Renewed Facility Operating Licenses DPR-31 and DPR-41 License Conditions 3.D state:
    "FPL shall implement and maintainin effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report (UFSAR) for Turkey Point Units 3 and 4 and as approved in the Safety Evaluation Report (SER) dated March 21, 1979 and supplemented by NRC letters dated April 3, 1980, July 9, 1980, December 8, 1980, January 26, 1981, May 10, 1982, March 27, 1984, April 16, 1984, August 12, 1987, and by Safety Evaluations dated February 25, 1994, February24, 1998, October 8, 1998, December 22, 1998, May 4, 1999, and May 5, 1999, subject to the following provision:
The licensee may make changes to the approved Fire ProtectionProgram without priorapproval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire."
2.2     NRC Acceptance of the Fire Protection Licensing Basis In response to the NRC's May 11, 1976 request, FPL performed a fire hazards analysis which analyzed the PTN fire protection program against the guidance of Appendix A to Branch Technical Position (BTP) Auxiliary Power Conversion Systems Branch (APCSB) 9.5-1. The results of the analysis, in addition to proposed modifications and additions to the fire protection program, were communicated to the NRC by letters dated February 25, 1977, request for amendment dated December 22, 1977, and supplemental submittals dated June 5, June 23, July 14, August 28, and November 7, 1978. These submittals served as the basis for the Appendix A to BTP APCSB 9.5-1 Safety Evaluation, dated March 21, 1979, and the associated License Amendment No. 45 and No. 37 to PTN Units 3 and 4 respectively, which added a license condition for the completion of fire protection modifications and implementation of administrative controls.
On April 19, 1982, the NRC issued License Amendments No. 84 and No. 78 to PTN Units 3 and 4, respectively. The amendments consisted of changes to the Technical Specifications.
In a safety evaluation dated April 16, 1984 the NRC concluded that the PTN alternate shutdown design will provide one train of systems necessary to achieve and maintain safe shutdown conditions in the event of a fire in the control room, cable spreading room or the north-south breezeway and thus will meet the requirements of 10 CFR 50, Appendix R, Sections III.G.3 and III.L with respect to safe shutdown.
Page 4 Revision 00                                                                               Page 4

Florida Power & Light                           2.0 Overview of Existing Fire Protection Program In addition to the approval of the PTN alternative shutdown design, the NRC granted the following licensing actions:
     " LA-01-19840327, 10 CFR 50, Appendix R Exemption from Fixed Fire Suppression in the Control Room (III.G.3 Criteria),
     " LA-02-19840327, 10 CFR 50, Appendix R Exemption from Having an Oil Collection Tank Sized to Contain the Entire Lube Oil System Inventory (111.0 Criteria),
     " LA-03-19840327, 10 CFR 50, Appendix R Exemption from Providing 3-Hour Rated Fire Barriers Between Fire Zones 11, 12, 13, 14, 15 and 16 (III.G.2.a Criteria),
     " LA-04-19840327, 10 CFR 50, Appendix R Exemption from Providing a 3-Hour Rated Fire Door for the Charging Pump Rooms on the West Walls of Fire Areas N and 0 (lll.G.2.a Criteria),
     " LA-05-19840327, 10 CFR 50, Appendix R Exemption from Providing a Fire Suppression System for the Charging Pump Rooms Fire Areas N and 0 (lll.G.2.c Criteria),
     " LA-06-19840327, 10 CFR 50, Appendix R Exemption from Providing Emergency Lighting Units in the Units 3 and 4 Containment to Facilitate the Manual Operation of Accumulator Stop Valves and RHR Inlet Isolation Valves (lll.J Criteria),
     " LA-07-19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (llI.G.2.f Criteria),
     " LA-08-19840327, 10 CFR 50, Appendix R Exemption from Providing the Enclosure of One Charging Pump in a 1-Hour Barrier for Fire Areas N and 0 (llI.G.2.c Criteria),
     " LA-09-19840327, 10 CFR 50, Appendix R Exemption from Total Enclosure of One Diesel Generator Radiator Room By a 3-Hour Rated Fire Barrier (llI.G.2.a Criteria),
     " LA-10-19870812, 10 CFR 50, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (lll.G.2.d Criteria),
     " LA-11-19870812, 10 CFR 50, Appendix R Exemption from Detection and Automatic Fire Suppression Systems in Outdoor Fire Areas 76, 77, 78, 83, 87, and 90 which are Separated by a Horizontal Distance of 20 Feet or Greater (lll.G.2.b Criteria),
     " LA-12-19870812, 10 CFR 50, Appendix R Exemption from Providing Detection and Automatic Fire Suppression in Outdoor Fire Zones 76, 77, 78, 83, 87, and 90 which are Enclosed in a Fire Barrier Having a 1-Hour Rating (Ill.G.2.c Criteria),
     " LA-13-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area AAA from Fire Area A (IllI.G.2.a Criteria),
Page 5 Revision 00                                                                                Page 5

Florida Power & Light                          2.0 Overview of Existing Fire Protection Program
    "   LA-14-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area F from Fire Area A (IIl.G.2.a Criteria),
    "   LA-15-19980224, 10 CFR 50, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier Until a Horizontal Distance of 10 Feet is Attained in Roof Top Locations Excluding the Turbine Building (llI.G.2.a Criteria),
    "   LA-16-19980224, 10 CFR 50, Appendix R Exemption from III.G.2.a for 3 Hr Electrical Raceway Fire Barriers in the Outdoor Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120 Excluding the Turbine Building (llI.G.2.a Criteria)
    "   LA-17-19981008, 10 CFR 50, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier until a Horizontal Distance of 20 Feet is Attained in the Outdoor Fire Zones 79-partial, 84-partial, 86, 88-partial, and 89-partial Excluding the Turbine Bldg (Ill.G.2.a Criteria),
    "   LA-18-19981008, 10 CFR 50, Appendix R Exemption from IIl.G.2.a West of the Open Turbine Building Structure Column Line A by a 1-Hour Rated Fire Barrier Until a Horizontal Distance of 20 Feet is Attained in Fire Zones 81 and 86 (lll.G.2.a Criteria),
    "   LA-19-19990504, 10 CFR 50, Appendix R Exemption from Technical Requirements for Outdoor Fire Zone 106R to the Extent that it had not Protected a Redundant Train with a 3-Hour Rated Barrier (lll.G.2.a Criteria),
    "   LA-20-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using a Minimum 1-Hour Barrier between Column Lines A and E-1(III.G.2.a Criteria),
    "   LA-21-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using Separation by a 25-Minute Fire Rated Barrier between Column Lines E-1 and Jc and the Turbine Deck (llI.G.2.a Criteria), and
    "   LA-22-20060927, 10 CFR 50, Appendix R Exemption for Lack of Suppression in the Mechanical Equipment Room (Fire Zone 097) and for Lack of Suppression and Detection in the Control Room Roof (Fire Zone 106R) (III.G.3 Criteria).
On February 25, 1994, the NRC issued License Amendment No. 159 and 153 to PTN Units 3 and 4 respectively. These amendments relocated the fire protection requirements from the technical specifications to the FSAR and revised the license condition 3.G and 3.F for PTN Units 3 and 4, respectively. (Note: In the Renewed Facility Operating Licenses fire protection is condition 3.D for both units).
Page 6 Revision 00                                                                               Page 6

Florida Power & Light                                                  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, FPL 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.
Page 7 Revision 00                                                                            Page 7

Florida Power & Light                                                                                                                           3.0 Transition Process Establish fundamental Ore         NFPA 805 Section 2.2(a) prutectiun elements tChapter 3)
Identity fire hazards        NFPA 808 Section 2.21b)
Nuclear safety NFPA 805 Section 2.2(c)              I         examined1)
(Chapter                                Property damage/business Interruption Evaluate compliance to                                                                                Radiation release performance criteria Identify stdmcttres, systems, or componrents (SSCS) in each fire         NFPA 805 Section 2.2(d) area to which 'he performance chite ria applies licese 50pprachpp.R, asi~etrmiistc(1                805 Section 2.2(e) pprvedperformiaince-BasediApproach                                    nNFPA CF Maintain compliance with existing plant                                                Evaluate ability  to satisfy performance requirements Exemptions, Engineering Evaluations)                                                                  (Chapter 4)
Determin          Be                                                        Define fire scenarios and fire design basis reuiem nt arte Veiydtri istic                                   qiaec                          o, each fire area being considered.
Evatuations                             Evaluate using, e.g.,
Fire modeling to quantify the fire isk 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-infoumed Change Evaluation Evalu ate risk impact of changes to NFPA 805 Section 2.2(h) the approved design basis 348sExamples 8e                  nt                 Design Basis Documents Fire haoards analysis (l o   m gandcic nRI-P ts t BtNew hAorransnigtion c
met         n hodnbascontrol                           i                Nuclear safety capability assessment n N                  Supporting engineerng calculations 2Probabilistic                                         safety analysis Risk-informed change evaluations Feedback                Establish monitoring program F       NFPA IM5Section 2.2(i)
Figure 3-1 NFPA 805 Process [NEI 04-02 Figure 3-1 based on Figure 2-2 of NFPA 805]2 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 (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.
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.
Revision 0                                                                                                                                                       rage a rage 8

Florida Power & Light                                                               3.0 Transition Process 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 A
Report
                                                                                             }   Sect. 4.5 Transition Figure 3-2 Transition Process (Simplified) [based on NEI 04-02 Figure 4-1]
3.4     NFPA 805 Frequently Asked Questions (FAQs)
The NRC has worked with NEI and two Pilot Plants (Oconee Nuclear Station and Harris Nuclear Plant) 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 Issues Summary (RIS) 2007-19, Process for Communicating Clarifications Page 9 Revision 0 Revision   0                                                                                           Page 9

Florida Power & Light                                                3.0 Transition Process 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.
Page 10 Revision 00                                                                        Page 10

Florida Power & Light                              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 PTN fire protection program to the requirements of NFPA 805 Chapter 3 was performed and documented in the report entitled "NFPA 805 Chapter 3 Compliance Review." The report 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) were 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 via 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 via Engineering Evaluations - For those sections/subsections determined to be equivalent to the NFPA 805 Chapter 3 requirements as documented by engineering analysis, and
     " Submit for NRC Approval/ License Amendment Required - 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.
Page 11 Revision 00                                                                           Page 11

Florida Power & Light                                                                                                                                                                            4.0 Compliance with NFPA 805 Requirements ExuaingFundamentalFireP*otection ProgramandDesignEtemerde nteIntrCompliance osas InReferenceDocumentFieldproide:
Meant            Yes tineNFPA8D5ChNpter 3 Requirement Approval                                                       that n...r JnComplioec          Inper3Dv       StatPement            clrctio                                  Donumeet            iuRde:u      ItemstoumnddurngRpeu d         powrvlpod           necrts       fl~~esblbl           eadn      h s*frwihpeA*ap~tI                    e NopsFud                               tNt                      a teo i e:U.b                 Reference InDocume           n Field thheatt References Etanoceten*  Fronds etailsondenronstiat                       derrfieldt complianceDcmatnyOe                                             ifneesnry Reid-NFPAOS            Yakeomgrotai           Ente                                                    onrine~oeantaipvr                                                       Note1ftttIM...
cepteb-chnproidedt                                 cntaorsof lnIineoertpprocatdoasrrar           ersdelaitsotWnpreoil-pter   n    n                               pe                                                                                                                              pprmalprode an e. ptis)of ticense,ubmit           regardingte issueforatish pre R, eppr at is being ciarmd. Ftecenfrelercrpi     anne ubm        bereti at&#xfd;      acerpt of inoarmaedpproedi t n ineC pace   B     asi C innSid                       e PcoN.- ypro ab                 L             ..                                                                                                                            fiete.it necessry.
CopaneSttmet              Ctm~arttat     emenetmT                                                        overtduringReRrieI Endsr ombesrint                      ComptianceBassInFate         t        nceDoumentFieldp       iote                                        2. E tg    Errtneenng    EqiureincpEvaluations  (prenusly Icorn as Generac  Lntter86-1Se luations. ecemdptiond.
Eea                      .s Etng Engineertng            provie                             DcumentRefrece that provid:                                    rr                          .                                   fe p e Engineeri n                         uaeny       uaRuon              S    um eof bases.for d                 de          compli onstrafte                              DocumentenyOpen       ed       deciane      performed from       for   prote      dig cn     e    qireee tsS ec      t ere dr        andfira rter
                                            . Ii-             -    F       nd    a nm e    talFin   re                                                   _a fnd dErig           n           E     clearlydemntrs n quivalet              of fireprotecti compared  tte      te rnsti reqsuirents to Evalit.?
Feldbe                                                                                                                           transditned."
C3tptenc o                                Enter               pronide:efoCorrecrnet                              Actions.a appropriate                         r putnt            Note       er ActionReiuated indoatee en interimpoedion  usedduring inde         drntMe           Feuther 4.2.Action.oof Required                Actin to be                   DocumenRefrece                                  Document    anyOpen         tie process ofcm etingtie B-1Table. FurtherAcmeRequiedentries Equsaencyin Cmpt~nce tateent                         nginerin supa C             luatent                                                    ItemsfounddurindRadntcmpne                           linne sciofRe itmsfondduin tdevitos fromn       Rcio      poew io true    eemnt btao           eormne d priortosuemnutingeto tin LAR... Ifthny fSR         toe arenetthinan n~quaal            Fildavntltnnator-y                                                                                                         actiudtyshouldbeaddedtotineLARsubmittal Not 3In                            omliaceBas on isNEI   l                Re02fegren Douet4il-2]3
                                    .h-.                                                    Stteen Eantimterrc                  eree .. d toopiacnne                         I  eemneDouetntl pons Figure 4-1             -  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 inSection 4.2.2.
Revision 0                                                                                                                                                                                                                                                                               Page 12

Florida Power & Light                              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 PTN 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, "Previouslyapproved alternativesfrom the fundamentalprotection program attributes of this chapter by the authority having jurisdiction(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.
PTN requests that the NRC concur with their finding of prior approval for the following sections of NFPA 805 Chapter 3:
    -   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, and therefore will be submitted to the NRC as "License Amendment Required":
0   3.3.3 - Approval is requested for the use of epoxy floor coatings as compliant with NFPA 805 Section 3.3.3, 0   3.3.5.1 - Approval is requested for the use of non-listed video/communication/data cables routed above suspended ceilings, 0   3.3.5.2 - Approval is requested for the use of PVC/HDPE conduits in embedded installation, 0   3.3.12(1) -Approval is requested for the potential of oil misting from the reactor coolant pumps due to normal motor consumption not captured by the oil collection system designed for pressurized and non-pressurized leakage and spillage, 0   3.5.5 and 3.9.4 - Approval is requested for lack of fire rated separation between fire pumps and lack of automatic suppression protection over the diesel-driven fire pump, 0   3.8.1(2) and 3.9.3 - Approval is requested for the lack of alarms on automatic water-based suppression systems to a constantly attended location, and Revision 0                                                                          Page 13

Florida Power & Light                              4.0 Compliance with NFPA 805 Requirements
     = 3.3.8 and NFPA30 Sections 2-1.1.1, 2-2.4.2, 2-4.4.3, and 2-5.2 -Approval is requested for deviations from NFPA 30 The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements are provided in Attachment L. PTN requests NRC approval of these performance-based methods.
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 Containments, Auxiliary Building, Control Building, Fuel Buildings, EDG Buildings, Radioactive Waste Processing, Raw Water Storage and Treatment, Turbine Building, and Intake Structure, and stand-alone structures or equipment that are identified in the facility's pre-transition licensing basis.
The structures in the protected area were evaluated to determine those that contain equipment that is required to meet the nuclear safety and radioactive release criteria described in Section 1.5 of NFPA 805. The protected area was chosen as the starting point for Turkey Point Units 3 and 4, since the area outside the protected area, but inside the owner controlled area, contains no structures that could potentially affect the nuclear safety and radioactive release criteria. The area outside the protected area contains structures such as parking lots, training facilities, fossil plants, switchyard, recreation facilities, natural areas, etc. None of those structures affect nuclear plant operations and none contain radioactive materials with credible release potential.
For structures within the protected area, each structure was examined for the potential to affect power plant operation, affect the ability to achieve and maintain the fuel in a safe and stable condition in the event of a fire, or contain radioactive materials that could be released in the event of a fire.
These structures are listed in Attachment I and define the "power block" and "plant".
Any future changes to this listing are subject to the configuration management controls discussed in Section 4.7.2.
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:
a   Establishing compliance with NFPA 805 Section 2.4.2, Revision 0                                                                             Page 14

Florida Power & Light                                4.0 Compliance with NFPA 805 Requirements
    "   Establishing the Safe and Stable Conditions for the Plant,
    " Establishing Recovery Actions, and
    " 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.
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 nuclear safety capability assessment. The following steps shall be performed:
(1) Selection of systems and equipment and their interrelationshipsnecessary to achieve the nuclearsafety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclearsafety performance criteria in Chapter 1 (3) Identification of the location of nuclearsafety 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 existing post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 1 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, or 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 Revision 0                                                                             Page 15

Florida Power & Liqht                            4.0 Compliance with NFPA 805 Requirements consequences, these sections of NEI 00-01 can be dispositioned without further review.
The comparison of the PTN post-fire SSA to NEI 00-01 Chapter 3 (NEI 04-02 Table B-
: 2) was performed and documented in the report entitled "NFPA 805 NSCA Methodology Review - NEI 04-02 Table B-2."
In addition, a review of NEI 00-01, Revision 2, Chapter 3 (ML091770265), was conducted against the following substantive changes applicable to an NFPA 805 fire protection program in the guidance from NEI 00-01, Revision 1.
    " Post fire manual operation of rising stem valves in the fire area of concern (NEI 00-01 Section 3.2.1.2),
    " Analysis of open circuits on a high voltage (e.g., 4.16 kV) ammeter current transformers (NEI 00-01 Section 3.5.2.1), and
    " Analysis of control power for switchgear with respect to breaker coordination (NEI 00-01 Section 3.5.2.4).
This review was performed and documented in the report entitled "NFPA 805 NSCA Methodology Review - NEI 04-02 Table B-2."
Results from Evaluation Process The method used to perform the 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 directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B with the following exception:
* Committed Modification: Perform modifications as necessary on DP412A, 4DP86 and 4DP87 to assure all power supplies have adequate breaker coordination.
See Committed Modification in Table S-2 of Attachment S.
Page 16 Revision 00 Revision                                                                            Page 16

Florida Power & Light                                                        4.0 Compliance with NFPA 805 Requirements Florida Power & Light                                                        4.0 Compliance with NFPA 805 Requirements S  I 1p       {                                Assemble Documentaeion o      2Determine                               and Document St    P2   {                                  Applicability of NEI00-01 Sections AppCable NE            nFor e     i--}
CSompa*rn         vts.0-*0-1evsi NEI Figure Summary 4-2 of Nuclearonsaft LmethodprioloyRvenrcs                          FQ0-09 Ants adtoed of NEI 00-01,f                   Reisison    2 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)
The Recovery Action Feasibility Evaluation determined that, in fire areas where operation of rising stem manual valves were used to achieve nuclear safety performance criteria, that fire damage in that area would not prevent the valve's operation because of separation (either by fire barrier or by distance) between the valve to be operated and the cables which failed a redundant shutdown path.
* Analysis of open circuits on a high voltage (e.g., 4.16 kV) ammeter current transformers (NEI 00-01 Section 3.5.2.1)
Evaluation of current transformers in the power block determined that none of the plant's current transformers would result in secondary damage because of an open circuit.
uAnalysis of control power fr switchgear with respect to breaker coordination (NEI 00-01 Section 3.5.2.4)
Evaluation of plant switchgear requiring control power to provide overcurrent protection determined several instances where the potential for fire-induced loss of breaker coordination could occur. Where these instances occurred, either modifications (either to protect cables or to prevent the cable damage from affecting the overcurrent trip of the circuit) are being performed to eliminate the Page 17 Revision 00                                                                                                     Page 17

Florida Power & Light                              4.0 Compliance with NFPA 805 Requirements concern or a fire risk evaluation determined that the risk of the fire-induced loss of breaker coordination was not significant.
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 pressurized water reactor. For all other configurations, safe and stable conditions are 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 maintaining the 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.
Results Based on the report entitled "Primary Control Station and Safe & Stable Plant Mode" the NFPA 805 licensing basis for PTN for a safe and stable condition in the event of a fire starting with the reactor in Mode 1, 2, or 3 is to maintain safe and stable conditions in Mode 3 with the ability to cool down and place residual heat removal (RHR) system in service if necessary. The final end state for safe and stable will be determined by the extent of equipment damage, existing inventory in the condensate storage tank, and the ability to provide makeup to the condensate storage tank. If any of these conditions indicate it is advisable to place RHR system in service then the plant will have the capability to do so. As a result, the equipment required to initiate and maintain hot shutdown cooling including the initiation of RHR cooling remains part of the at-power analysis. Initiation of RHR cooling does not imply however, that the plant would proceed all the way to cold shutdown. The plant may remain on RHR cooling at other than cold shutdown conditions for an extended period of time. The ability to maintain this condition for extended periods will require additional actions such as replenishment of diesel fuel oil, etc. These actions are already part of the emergency response procedures. These are considered to be low risk (not adding to the total fire risk) due to being proceduralized and the time and resources required to perform such actions.
Revision 0                                                                             Page 18

.Florida Power & LiQht                                4.0 Compliance with NFPA 805 Requirements Thus, the Nuclear Safety Performance Criteria needs to demonstrate:
     " The ability to maintain the nuclear fuel in a safe and stable condition in Mode 3, and
     " The ability to proceed to RHR cooling if so chosen given the plant configuration and equipment availability following a fire.
This is performed in two analyses.
     " At-Power analysis for fires originating in Modes 1 - 3. This analysis is discussed in Section 4.2.4.
     " Non-Power Analysis for potential fires originating in Modes 4 and lower. This analysis is discussed in Section 4.3.
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 (Regulatory Position 2.21 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) (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 documents entitled, NFPA 805 Fire Area Transition Review, Turkey Point Nuclear Power Plant Fire Risk Evaluations, and NFPA 805 Recovery Action Feasibility Evaluation. Refer to Attachment G for the detailed evaluation process and summary of the results from the process and Attachment W for additional information on the risk-informed treatment of recovery actions.
Page 19 Revision 00                                                                             Page 19

Florida Power & Light                            4.0 Compliance with NFPA 806 Requirements 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 multiple spurious operations (MSOs) for NRC review and approval. As a minimum, NEI 04-02 suggests 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 utilized to address MSOs for PTN is summarized below.
As part of the NFPA 805 transition project, a review and evaluation of PTN 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 Pressurized Water Reactor Owners Group (PWROG) generic MSO list included in NEI 00-01, Revision 2, dated June 5, 2009 was utilized.
The approach outlined in Figure 4-3 (based on Figure 4-8 from FAQ 07-0038) is one acceptable method to address fire-induced MSOs. 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, Revision 2 Section F.4.2),
* Updating the Fire PRA model and the NSCA include the MSOs of concern,
* Evaluating for NFPA 805 Compliance, and
* 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 may be part of the plant change review process and/or inspection process).
Revision 0                                                                           Page 20

Florida Power & Light                                                 4.0 Compliance with NFPA 805 Requirements Step    -             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 Ev/a       for NFPA 805           No Step 4                       Compli       e                   Pursue other resolution options Compliant with NFPA 805?
StepYes 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 and the results.
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 (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:
0 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, Page 21 Revision 00                                                                                                    Page 21

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements
    "   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 the report entitled, NFPA 805 Transition Review of Existing Engineering Equivalency Evaluations.
In accordance with the guidance provided 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 used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in the Attachments A and C as appropriate.
In addition, none of the transitioning EEEEs require NRC approval.
4.2.3   Licensing Action Transition Overview of Evaluation Process The existing licensing actions (exemptions and safety evaluations) 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, and
    " Determination that these bases for acceptability are still valid and required for NFPA 805.
Results Attachment K contains the detailed results of the Licensing Action Review.
The following licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved (NFPA 805 Section 2.2.7). These licensing actions are considered compliant under 10 CFR 50.48(c).
Page 22 Revision 00                                                                          Page 22

Florida Power & Light                               4.0 Compliance with NFPA 805 Requirements
    "   LA-02-19840327, 10 CFR 50, Appendix R Exemption from Having an Oil Collection Tank Sized to Contain the Entire Lube Oil System Inventory (111.0 Criteria),
    " LA-07-19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lII.G.2.f Criteria), and
    " LA-10-19870812, 10 CFR 50, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (IIl.G.2.d Criteria).
The following licensing actions are no longer necessary and will not be transitioned into the NFPA 805 fire protection program:
    " LA-01-19840327, 10 CFR 50, Appendix R Exemption from Fixed Fire Suppression in the Control Room (III.G.3 Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-03-19840327, 10 CFR 50, Appendix R Exemption from Providing 3-Hour Rated Fire Barriers Between Fire Zones 11, 12, 13, 14, 15 and 16 (llI.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-04-19840327, 10 CFR 50, Appendix R Exemption from Providing a 3-Hour Rated Fire Door for the Charging Pump Rooms on the West Walls of Fire Areas N and 0 (lll.G.2.a Criteria)
This exemption is no longer required because the subject boundaries have been demonstrated adequate for the hazard in an existing engineering equivalency evaluation.
    " LA-05-19840327, 10 CFR 50, Appendix R Exemption from Providing a Fire Suppression System for the Charging Pump Rooms Fire Areas N and 0 (lIl.G.2.c Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-06-19840327, 10 CFR 50, Appendix R Exemption from Providing Emergency Lighting Units in the Units 3 and 4 Containment to Facilitate the Manual Operation of Accumulator Stop Valves and RHR Inlet Isolation Valves (lIl.J Criteria)
This exemption is no longer required because NFPA 805 does not require 8 hour battery backed emergency lights.
    " LA-08-19840327, 10 CFR 50, Appendix R Exemption from Providing the Enclosure of One Charging Pump in a 1-Hour Barrier for Fire Areas N and 0 (lll.G.2.c Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
Revision 0                                                                            Page 23

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements
    "   LA-09-19840327, 10 CFR 50, Appendix R Exemption from Total Enclosure of One Diesel Generator Radiator Room By a 3-Hour Rated Fire Barrier (lII.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    "   LA-11-19870812, 10 CFR 50, Appendix R Exemption from Detection and Automatic Fire Suppression Systems in Outdoor Fire Areas 76, 77, 78, 83, 87, and 90 which are Separated by a Horizontal Distance of 20 Feet or Greater (lll.G.2.b Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    "   LA-12-19870812, 10 CFR 50, Appendix R Exemption from Providing Detection and Automatic Fire Suppression in Outdoor Fire Zones 76, 77, 78, 83, 87, and 90 which are Enclosed in a Fire Barrier Having a 1-Hour Rating (lll.G.2.c Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    "   LA-13-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area AAA from Fire Area A (lll.G.2.a Criteria)
This exemption is no longer required because the subject boundaries have been demonstrated adequate for the hazard in an existing engineering equivalency evaluation.
    "   LA-14-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area F from Fire Area A (llI.G.2.a Criteria)
This exemption is no longer required because the subject boundaries have been demonstrated adequate for the hazard in an existing engineering equivalency evaluation.
    "   LA-15-19980224, 10 CFR 50, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier Until a Horizontal Distance of 10 Feet is Attained in Roof Top Locations Excluding the Turbine Building (lll.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    "   LA-16-19980224, 10 CFR 50, Appendix R Exemption from IIl.G.2.a for 3-Hour Electrical Raceway Fire Barriers in the Outdoor Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120 Excluding the Turbine Building (llI.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    "   LA-17-19981008, 10 CFR 50, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier until a Horizontal Distance of 20 Feet is Attained in the Revision 0                                                                           Page 24

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements Outdoor Fire Zones 79-partial, 84-partial, 86, 88-partial, and 89-partial Excluding the Turbine Bldg (lll.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-18-19981008, 10 CFR 50, Appendix R Exemption from IIl.G.2.a West of the Open Turbine Building Structure Column Line A by a 1-Hour Rated Fire Barrier Until a Horizontal Distance of 20 Feet is Attained in Fire Zones 81 and 86 (IllI.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-19-19990504, 10 CFR 50, Appendix R Exemption from Technical Requirements for Outdoor Fire Zone 106R to the Extent that it had not Protected a Redundant Train with a 3-Hour Rated Barrier (lll.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-20-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using a Minimum 1-Hour Barrier between Column Lines A and E-1(lII.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-21-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using Separation by a 25-Minute Fire Rated Barrier between Column Lines E-1 and Jc and the Turbine Deck (llI.G.2.a Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
    " LA-22-20060927, 10 CFR 50, Appendix R Exemption for Lack of Suppression in the Mechanical Equipment Room (Fire Zone 097) and for Lack of Suppression and Detection in the Control Room Roof (Fire Zone 106R) (III.G.3 Criteria)
This exemption is no longer required because the fire risk evaluation has found that the fire area is compliant with NFPA 805 Section 4.2.4.
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), PTN requests that the exemptions listed in Attachment K be rescinded as part of the LAR process.
See Attachment 0, Orders and Exemptions. It is PTN's understanding that implicit in the superseding of the current license condition, all prior fire protection program safety evaluation reports and commitments will be superseded in their entirety.
Page 25 Revision 00 Revision                                                                              Page 25

Florida Power & Light                               4.0 Compliance with NFPA 805 Requirements 4.2.4 Fire Area Transition Overview of Evaluation Process The Fire Area Transition (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 document the results of the review. See Section 4.2.3.
    " 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 (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 Attachment C (NEI 04-02 Table B-3).
    " 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 (including fire area licensing actions and engineering evaluations) and the NFPA 805 Section 4.2.4 compliance strategies (including simplifying deterministic assumptions) to determine the scope of fire protection systems and features 'required' by NFPA 805 Chapter 4. The 'required' fire Revision 0                                                                               Page 26

Florida Power & Light                           4.0 Compliance with NFPA 805 Requirements protection systems and features are subject to the applicable requirements of NFPA 805 Chapter 3.
Figure 4 Summary of Fire Area Review
[Based on FAQ 07-0054 Revision 1]
Page 27 Revision 00                                                                       Page 27

Florida Power & Light                               4.0 Compliance with NFPA 805 Requirements Results of the Evaluation Process Attachment C contains the results of the Fire Area Transition review (NEI 04-02 Table B-3). On a fire area basis, Attachment C, Table C-1 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 - Specific references to exemption requests that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability of the licensing action should be provided. Attachment T contains items for which PTN is requesting concurrence of prior approval.
     " 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.
     " Summaries of Fire Risk Evaluations performed for variances from the deterministic requirements are also 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.
4.3     Non-Power Operational Modes 4.3.1 Overview of Evaluation Process PTN implemented the process outlined in NEI 04-02 and FAQ 07-0040, Non-Power Operations Clarification. The goal (as depicted in Figure 4-5) 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, Page 28 Revision 00 Revision                                                                                Page 28

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements
    "   Performed Fire Area Assessments (identify pinch points - plant locations where a single fire may damage all success paths of a KSF), and 0
    " Managed 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.
Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points Page 29 Revision 00 Revision                                                                                Page 29

Florida Power & Liciht                                            4.0 Compliance with NFPA 805 Requirements Flrd oe           ih                                        40CmlaewithNP            85    aireet Higher Risk Evolution as Defined by Plant Specific       Higher Risk       No Outage Risk Criteria for example                          Evolution?
: 1) Time to Boil
: 2) Reactor Coolant System and Fuel Pool Inventory
: 3) Decay Heat Removal Yes Determine Fire Area Impact based on NPO Fire Area Assessments KSFNo-
                                                        <    Lost?         No Yes Implement Contingency Plan for Specific KSF Figure 4-6 Manage Pinch Points Page 30 Revision 00 Revision                                                                                             Page 30

Florida Power & Light                               4.0 Compliance with NFPA 805 Requirements 4.3.2 Results of the Evaluation Process Based on FAQ 07-0040, the Plant Operating States considered for equipment and cable selection are defined in the report entitled, NFPA 805 Non-Power Operations Review.
For each unit, components were identified to provide the KSFs of Decay Heat Removal, Electrical Distribution System, Inventory and Reactivity Control, and Shutdown Monitoring Instrumentation. The selection of equipment was further broken down based on consideration of KSF success paths. The selected components were flagged as NPO within the NICAD database to allow 'pinch point' analysis within an analysis area.
For those components not already in the NICAD database or which had a different functional state for non-power operations from that in the safe shutdown analysis, cables selection and routing was performed as described in the plant's Nuclear Safety methodology. The resulting information was entered into the NICAD database and a series of NPO reports were developed within the software to allow the 'pinch point' analysis to be performed.
The report contains the fire zone assessments of the 'pinch point' analysis and recommendations of changes to outage management procedures and other administrative controls. In accordance with FAQ 07-0040, any analysis area in which all of the credited success paths for a given KSF are lost is considered a 'pinch point.'
The list of recommendations specified in the evaluation considers the following actions from FAQ 07-0040:
    " Restriction of hot work in the analysis areas during periods of increased vulnerability,
    " Verification of functional detection and / or suppression in the vulnerable analysis areas,
    " Limitation of transient combustible materials in the analysis areas during periods of increased vulnerability,
    " Plant equipment configuration changes (e.g., removing power from equipment once it is placed in its desired position),
     " Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during periods of increased vulnerability, and
     " Reschedule the work to a period with lower risk or higher defense-in-depth.
See Attachment D for more complete details. Based on incorporation of the recommendations from the 'pinch point' evaluations into appropriate plant procedures, prior to implementation of NFPA 805, the performance goals for Non-Power Operations are fulfilled and the requirements of NFPA 805 are met. See Implementation Item 14 in Table S-3 of Attachment S.
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 Page 31 Revision 0 Revision  0                                                                           Page 31

Florida Power & Light                                4.0 Compliance with NFPA 805 Requirements contained in the report entitled, PTN NFPA 805 Radioactive Release Review. 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, etc.) 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.
     " A review of engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents). This review included all plant operating modes (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 The radioactive release review determined the fire protection program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the implementation items identified in Attachment E. See Implementation Item 15 in Table S-3 of Attachment S.
The main strategy for complying with the radioactive release requirements is ensuring that all buildings or areas containing radioactive hazards or the potential for an uncontrolled release during a fire have adequate strategies to minimize the uncontrolled release of radioactive material during fire fighting activities. This includes the revision or creation of documentation such as pre-fire plans, fire brigade training materials, standard operating procedures, and administrative controls.
If the engineering controls (ventilation, drainage) identified in Attachment E were to be lost in the event of a failure then training, pre-fire plans and standard operating procedures will compensate for the loss.
The site specific review of the direct effects of fire suppression activities on radioactive release is summarized in Attachment E.
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 (discussed in Section 4.5.1 and Attachments U, V, and W).
     " NFPA 805 Performance-Based Approaches (discussed in Section 4.5.2).
4.5.1    Fire PRA Development and Assessment In accordance with the guidance in RG 1.205, a Fire PRA model was developed for PTN 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 Applications, (hereafter referred to as Fire PRA Standard). PTN conducted a peer review by independent industry analysts in accordance with RG 1.200 Revision 0                                                                              Page 32

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements 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 PTN base internal events PRA (PTN PSA Model - EPU, PTN-BFJR-09-021, Revision 1) 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, FAQs, and the following model development documents:
     " Turkey Point Units 3 & 4, Fire Probabilistic Risk Assessment, Component and Cable Selection Report,
     " Turkey Point Units 3 & 4, Fire Scenario Report, NUREG/CR-6850 Tasks 8 and 11,
* Turkey Point Units 3 & 4, FPRA Summary Report, NUREG/CR-6850 Task 16,
* Turkey Point Units 3 & 4, Human Failure Evaluation, and
* Fire Ignition Frequency Development FP&L NFPA 805 Transition.
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 Fire modeling was performed as part of the Fire PRA development (NFPA 805 Section 4.2.4.2). 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 acceptability of the use of these fire models is included in Attachment J.
4.5.1.3     Results of Fire PRA Peer Review The PTN Fire PRA was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4.
The results (i.e., Supporting Requirement capability assessments and Facts &
Observations (F&Os)) documented in the April 2010 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 Revision 0                                                                           Page 33

Florida Power & Light                                        4.0 Compliance with NFPA 805 Requirements 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 March 2012 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 core damage frequency (CDF) and large early release frequency (LERF) were derived using the guidance for Fire PRA development in NUREG/CR-6850, FAQs, and report entitled, FPRA Summary Report, and is useful in identifying the areas of the plant where fire risk is greatest. A review of each fire initiating event that individually represents 1% or greater 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 (NFPA 805 Section 4.2.4.1), and
    " Fire Risk Evaluation (NFPA 805 Section 4.2.4.2).
4.5.2.1      Fire Modeling Approach The fire modeling approach was not utilized for the transition.
4.5.2.2      Fire Risk Approach Overview of Evaluation Process The Fire Risk Evaluations were completed as part of the PTN 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 (Appendix 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 Page 34 Revision 00                                                                                            Page 34

Florida Power & Light                               4.0 Compliance with NFPA 805 Requirements NFPA 805. A Fire Risk Evaluation was performed for each fire area containing variances from the deterministic requirements of Section 4.2.3 of NFPA 805 (VFDRs).
0 If the 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 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, and 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, and o Fire area change in risk summary Step 3 - Reviewed the Acceptance Criteria 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, Page 35 Revision 00                                                                           Page 35

Florida Power& Light                                 4.0 Compliance with NFPA 805 Requirements
            -   Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage, and
            -   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 reviews were 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.
            - 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 FRE.
Page 36 Revision 00                                                                             Page 36

Florida Power & Light                                   4.0 Compliance with NFPA 805 Requirements Prepare for Fire Risk Evaluation Determine How to Model              Discuss and Document in the VFDR in the Fire PRA              Fire PRA and Fire Risk 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 37 Revision 00                                                                                  Page 37

Florida Power & Light                              4.0 Compliance with NFPA 805 Requirements Results of Evaluation Process Disposition of VFDRs The PTN 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, Table C-1.
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, Regulatory Position 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, Regulatory Position 2.2.4.2 states in part "The total increase or decrease in risk associatedwith the implementation of NFPA 805 for the overall plant should be calculatedby summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions). The total risk increase should be consistent with the acceptance guidelines 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 additionalrisk associated with previously approved recovery actions is greaterthan the acceptanceguidelines in Regulatory Guide 1.174, then the net change in total plant risk incurredby any proposed alternatives to the deterministic criteria in NFPA 805, Chapter4 (other than the previously approvedrecovery actions), should be risk neutral or represent a 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 monitoringprogram shall be establishedto ensure that the availabilityand reliabilityof the fire protection systems and features are maintainedand 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."
Page 38 Revision 00 Revision                                                                              Page 38

Florida Power & Light                                      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 Implementation Item 2 in Table S-3 of 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, and
* 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 PTN NFPA 805 Monitoring Program Engineering Evaluation 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, and
              -    Required by Chapter 3 of NFPA 805.
o Nuclear Safety Capability Assessment equipment 4
              -    Nuclear safety equipment,
              -    Fire PRA equipment, and
              -    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.
Revision 0                                                                                              Page 39

Florida Power & Liqht                              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/or 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 program activities and will be documented in the NFPA 805 Monitoring Program Engineering Evaluation.
: 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 NFPA 805 Monitoring Program Engineering Evaluation.
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).
Fire protections systems and features that meet or exceed the criteria identified above are considered High Safety Significant (HSS) will be included in the monitoring program contained in the site Maintenance Rule Program described in NAP-415, Maintenance Rule Program Administration. The remaining required fire protection systems and features will be monitored via the existing inspection and test program and in the existing system / program health program as described in 0-ADM-016.3, Fire Protection Impairments, ER-AA-201-2001 System and Program Health Reporting, and ER-AA-201-2002 System Performance Monitoring.
: 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 Revision 0                                                                           Page 40

Florida Power & Light                            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 included in the 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/or 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 (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:
    " Transient Combustible Control; Transient Exclusion Zones,
    " Hot Work Control; Administrative Controls,
    " Impairment and compensatory measures including program compliance and effectiveness, and
    " Fire Brigade Effectiveness 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 system and program health programs. Fire protection health reports, self-assessments, regulator and insurance company reports provide inputs to the monitoring program.
Phase 3 - Risk Target Value Determination Phase 3 establishes the target values for reliability and availability for the fire protection systems and features the met or exceeded the screening criteria and the HSS NSCA equipment established in Phase 2.
Page 41 Revision 00                                                                             Page 41

Florida Power & Light                                4.0 Compliance with NFPA 806 Requirements Target values for reliability and availability for the fire protection systems and features are established 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 (-2-3 operating cycles). In addition, the EPRI Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" (Reference 28) will be used as input for establishing reliability targets, action levels, and monitoring frequency.
See Implementation Item 1 in Table S-3 of Attachment S 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. The Monitoring Program failure criteria and action level targets will be documented in the NFPA 805 Monitoring Program Engineering Evaluation.
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 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.
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 PI-AA-205, Condition Evaluation and Corrective Action, 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 (e.g., at a frequency of approximately every two to three operating cycles and coordinated with the NRC Triennial Inspection), taking Revision 0                                                                              Page 42

Florida Power & Liaht                              4.0 Cornoliance with NFPA 805 Reouirements into account, where practical, industry wide operating experience. This will be conducted as part of other established assessment activities. 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 43 Revision 00                                                                            Page 43

Florida Power & Light                                                       4.0 Compliance with NFPA 805 Requirements Phase 2 - Screening High Safety                                                                                     Yes Significance of feature by compartment?
High Risk        Yes Significance?
Yes Yes Include in MNo sintenance Rule?
Yes NFPA 805 Specific                           Normal System &
Monitoring Process                            Program Health Establish targets for                      Monitoring Process or                           Use Maintenance Rule reliability/unavailability in                      Outage Risk                                    for Monitoring Phase 3                             Management for NPO Fully describe process used*
No 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, PTN has documented analyses to support compliance with 10 CFR 50.48(c). The analyses are being performed in accordance with FPL's processes for ensuring assumptions are clearly defined, that results are easily Page 44 Revision 00 Revision                                                                                                              Page 44

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements 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 will be created as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation. Appropriate cross references will be established to supporting documents as required by FPL processes. Figure 4-9 depicts the planned post-transition documentation and relationships. See Implementation Item 16 in Table S-3 of Attachment S.
Page 45 Revision 00                                                                          Page 45

Florida Power & Light                                                                      4.0 Compliance with NFPA 805 Requirements NFPA 805 DOCUMENTS NSCA Database NSEL Camp                             Cale PRA Equipment                        Non-Power and Data                    Equipment and Data
              ------ -NSCA CALCULATION.
Comp & Cable                   FA Assessment Method/Results                 Method/Results                                        Revised License Condition MSO and       OMA Treatents             S SSA      Drawings Treatments NSCA SUPPORTING INFO THCalculations                Manual Action Revised UFSAR SFeasibility SB2TableTable FIRE SAFETY ANALYSIS (DBD)
Coordination                 Plant DBDs that                          On a Fire Area Basis Calculations I                 Support NSCA                            - Fire Area Description MHIF                                                             *FHA Database information
                                                                                    -Nuclear Safety Performance Criteria Compliance Summary (NEI 04-02 B-3 Table
     .......................................       ........ ... . . . .               Results)
                                                                               -10.
                                                                                    *Non-Power Evaluation Results Summary Non-Power Mode NSCA Treatment                                           -Radioactive Release Summary On a Generic Basis
                                                                                    *B-1 Table Results Non-Power Operations Calculations
                                                                                    -Radioactive Release (Training)
                                                                                    -Monitoring Program NFPA 805 FIRE RISK EVALUATIONS Fire Risk Evaluation Calculation(s)
                                                                         -                       Fire PRA FHA DATABASE DATA Ignition Sources                 FP Systems and
              & Scenarios                     Features Data Inventory of                      B-I Table Hazards                      Detailed Data FHA SUPPORT DOCUMENTATION FP Systems Code Compliance                       FP Drawings Evaluations Bold text indicates new NFPA 805 documents FP System and                       Engineering Feature DBDs                       Equivalency Evaluations Radioactive Fire Pre-Plans                 Release Review Calculation Figure 4 NFPA 805 Planned Post-Transition Documents and Relationships Page 46 Revision 00 Revision                                                                                                                                           Page 46

Florida Power & Light                                       4.0 Compliance with NFPA 806 Requirements 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and                                 0 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 FPL 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, and
    " 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-Revision 0                                                                                           Page 47

Florida Power & Light                             4.0 Compliance with NFPA 805 Requirements 03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).
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 that may include fire modeling and risk assessment techniques. 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 fire modeling analyses and 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 (change in core damage frequency) and ALERF (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 48 Revision 00 Revision                                                                             Page 48

Florida  Power & Light                                        4.0 Compliance opine with ihNP NFPA 805 Requirements 05Rqieet Flrd oe          ih                                      4.
Defining the Change (5.3.2)                                                                            0 License         No       vtChp3r            Ye AmendmentAm                            rve Requestprvosya License Amendment          r Request NOT Required Preliminary Risk Screening (5.3.3)
Risk Evaluation (5.3.4)
PRA Capability Category Assessment Fire PRA Capability Category Assessment A
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 Page 49 Revision 00                                                                                      Page 49

Florida Power & Light                             4.0 ComDliance with NFPA 805 Requirements The PTN fire protection program configuration is defined by the program documentation. To the greatest extent possible, 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 PTN documents and databases that currently exist will be revised as necessary to reflect the new NFPA 805 licensing bases requirements.
Several NFPA 805 document types such as: NSCA Supporting Information, Non-Power Mode NSCA Treatment, etc., 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, or
* 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. See Implementation Item 16 in Table S-3 of Attachment S.
4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality FPL will maintain the existing Fire Protection Quality Assurance program.
During the transition to 10 CFR 50.48(c), FPL performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.
Revision 0                                                                            Page 50

Florida Power & Light                              4.0 Compliance with NFPA 805 Requirements 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 FPL 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. FPL 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 FPL 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 the Turkey Point FPRA Summary Report. In addition, sensitivity to uncertainty associated with specific Fire PRA parameters was quantitatively addressed in this report.
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 FPL 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.
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 FPL procedures that require independent review.
Page 51 Revision 00 Revision                                                                             Page 51

Florida Power & Light                            4.0 Compliance with NFPA 805 Requirements 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, FPL will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work.
See Implementation Item 16 in Table S-3 of Attachment S.
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 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, Table C-2, Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features. Table C-2 provides the following information:
     " 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 (Note: Compliance is determined on a Fire Area basis therefore a compliance basis is not provided for individual fire zones.),
Revision 0                                                                           Page 52

Florida Power & Light                              4.0 Compliance with NFPA 805 Requirements 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 information is provided on a fire zone basis. 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), and o N - Non-Power Operations Criteria: Systems/Features required to meet the nuclear safety performance criteria of NFPA 805 as identified by the NPO analysis During the implementation phase, the PTN procedures and processes will be updated to reflect the NFPA 805 required fire protection systems and features. See Implementation Item 16 in Table S-3 of Attachment S 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.
The Fire PRA model represents the as-built, as-operated and maintained plant as it will be configured at the completion of the transition to NFPA 805. In addition, the Fire PRA model includes the extended power uprate (EPU) changes that are currently being implemented. The Fire PRA model includes credit for modifications completed as result of the NFPA 805 transition effort as identified in Table S-1 of Attachment S and the planned implementation of the modifications identified in Table S-2 of Attachment S.
Following installation of modifications and the as-built installation details, additional refinements surrounding the modification may need to be incorporated into the Fire PRA model. However, these changes are not expected to be significant. See Implementation Item 18 in Table S-3 of Attachment S. No other significant plant changes are outstanding with respect to their inclusion in the Fire PRA model.
Revision 0                                                                             Page 53

Florida Power & Liqht                          4.0 Compliance with NFPA 805 Requirements 4.8.3 Supplemental Information - Other Licensee Specific Issues There are no PTN specific issues that warrant additional treatment in this section Page 54 Revision 00 Revision                                                                           Page 54

Florida Power & Light                                                 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(c) 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 PTN does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, General Design Criterion (GDC) 3, Fire Protection. The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086).
    "NFPA 805 does not supersede the requirements of GDC 3, 10 CFR 50.48(a), or 10 CFR 50.48(0. Those regulatory requirements continue 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 requiredto meet the Chapter 1 performance criteria through the methodology in Chapter 4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii)requirement to limit fire damage to SSCs important to safety so that the capability to safely shut down the plant is ensuredis satisfiedby meeting the performance criteria in Section 1.5.1 of NFPA 805. The Section 1.5.1 criteriainclude 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 required to achieve the nuclearsafety 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 applicable requirements of NFPA 805, Chapter3. Having identified the requiredfire protection systems and features, the licensee selects either a deterministicor performance-basedapproachto demonstrate that the performance criteria are satisfied. This process satisfies the GDC 3 requirementto design and locate SSCs importantto safety to minimize the probabilityand effects of fires and explosions."
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 contents of the existing fire protection program plan required by 10 CFR 50.48(a) are Revision 0                                                                             Page 55

Florida Power& Liaht                                                                   5.0 Regulatory   Evaluation Flrd oe       &    ih                                                             . eilaovEauto 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 10 CFR 50, 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 PTN fire protection program and applicable industry and PTN documents that address the topic.
10 CFR 50.48(a)
Table 5-1 10 CFR 50.48(a) - Applicability/Compliance Reference 10 CFR 50.48(a) Section(s)                         Applicability/Compliance 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.
Page 56 Revision 00                                                                                               Page 56

Florida Power & Light                                                                    5.0 Requiatory Evaluation Floid Power            50Li8ht)5.                                     RelulatorCmEvaluation Table 5-1 10 CFR 50.48(a)   -  Applicability/Compliance Reference 10 CFR 50.48(a) Section(s)                            Applicability/Compliance Reference (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                   See RM-AA-100, Records Management Program.
superseded.
(4) Each applicant for a design approval, design                   Not applicable. PTN is licensed under 10 CFR 50.
certification, or manufacturing license under part 52 of 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 Page 57 Revision 0                                                                                                        Page 57

Florida Power & Li-qht                                                                 5.0 Regulatory Evaluation 10 CFR 50.48(c)
Table 5-3 10 CFR 50.48(c) - Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)                                 Applicability/Compliance Reference (1) Approval of incorporationby reference. National Fire Protection Association      General Information. NFPA (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for         805 (2001 edition) is the Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805),        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. NFPA this section, references to NFPA 805 are to the 2001 Edition, with the           805 (2001 edition) is the following exceptions, modifications, and supplementation:                        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/BusinessInterruption Goal, 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                Feed and bleed is not utilized performance criteria of Sections 1.5.1(b) and (c), a high-pressure               as the sole fire-protected safe charging/injection pump coupled with the pressurizer power-operated relief       shutdown methodology.
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.                         and "complies via previous NRC approval" with application of a fire retardant cable coating in certain areas as documented in Attachment A.
(vi) Water supply and distribution. The italicized exception to Section 3.6.4 is PTN identified this section as not endorsed. Licensees who wish to use the exception to Section 3.6.4          N/A as documented in must submit a request for a license amendment in accordance with                 Attachment A.
paragraph (c)(2)(vii) of this section.
Revision 0                                                                                                     Page 58

Florida Power & Light                                                                  5.0 Regulatory Evaluation Table 5-3 10 CFR 50.48(c)-Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)                                Applicability/Compliance 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 &sect; 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 &sect; 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.
Page 59 Revision 00                                                                                                  Page 59

Florida Power & Liqht                                                                    5.0 Regulatory Evaluation Table 5-3 10 CFR 50.48(c)-Applicability/Compliance Reference 10 CFR 50.48(c) Section(s)                                 Applicability/Compliance Reference (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 &sect; 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.
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).
5.2        Regulatory Topics 5.2.1      License Condition Changes The current PTN fire protection license condition 3.D 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 FPL conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded. FPL determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the PTN adoption of the new fire protection licensing basis.
5.2.3 Orders and Exemptions A review was conducted of the PTN 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 .
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. FPL has evaluated the proposed amendment and determined that it involves no significant hazards consideration. 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:
Revision 0                                                                                                     Page 60

Florida Power & Light                                                5.0 Regulatory Evaluation
     "  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. FPL 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 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 PTN UFSAR After the approval of the LAR, in accordance with 10 CFR 50.71(e), the PTN UFSAR will be revised. The format and content will be consistent with NEI 04-02.
5.5     Transition Implementation Schedule The following schedule for transitioning PTN 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 unless that falls within a scheduled outage window. Then this will occur 60 days after startup from that scheduled outage.
See Attachment S Table S-3.
     " Modifications will be completed by the startup of the second refueling outage (for each unit) after the issuance of the SER. Appropriate compensatory measures will be maintained until modifications are complete. See Attachment S Table S-2.
Page 61 Revision 00                                                                           Page 61

Florida Power & Light                                                        6.0 References

==6.0 REFERENCES==

The following references were used in the development of the Transition Report.
Additional references are in the NEI 04-02 Tables in the various Attachments.
NRC Documents
: 1) Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, February 28, 1992
: 2) 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)
: 3) NRC Enforcement Policy, Policy Statement: Revision, Federal Register, Vol. 69, No. 115, June 16, 2004, pp. 33684-33685
: 4) 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
: 5) NRC Regulatory Issue Summary 2007-19: Communicating Clarifications of Staff Positions in RG 1.205 Concerning Issues Identified During Pilot Application of NFPA Standard 805, August 20, 2007 (ML0611660105)
: 6) NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, April 2005
: 7) 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
: 8) Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2
        - March 2009)
: 9) Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Revision 1, December 2009 10)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 1006756, Final Report, July 2003
: 3) EPRI Report 1010068, Aggregation of Quantitative Risk Assessment Results, December 2005
: 4) NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Revision 1, January 2005 Revision 0                                                                          Page 62

Florida Power & Light                                                    6.0 References
: 5) NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Revision 2, May 2009
: 6) NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c), Revision 2, April 2008
: 7) NEI 05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, Nuclear Energy Institute, Revision 2, November 2008
: 8) NEI 05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, Nuclear Energy Institute, Revision 1, Draft G, November 2007
: 9) NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Nuclear Energy Institute, Revision 0, November 2008 10)NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition 11)NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management Licensee Correspondence
: 1) FPL to NRC, Letter of Intent to Adopt NFPA 805, November 15, 2005 (ML053290175)
: 2) NRC to FPL, Period of Enforcement Discretion During Implementation of NFPA 805, September 18, 2006 (ML062020265)
: 3) FPL to NRC, Request for Extension of Enforcement Discretion and Revised to Submittal Schedule for 10 CFR 50.48(c) License Amendment Request, September 11, 2008 (ML082600482)
: 4) NRC to FPL, Evaluation of the Request for an Extension of the Enforcement Discretion During Transition to NFPA 805, November 14, 2008 (ML083180165)
: 5) FPL to NRC, Request for Extension of Enforcement Discretion and Commitment to Submittal Dates for 10 CFR 50.48(c) License Amendment Request, June 23, 2011 (ML11178A237)
: 6) NRC to FPL, Commitment to Submit a License Amendment Request to Transition to 10 CFR 50.48(c) and Request to Extend Enforcement Discretion, August 8, 2011 (ML11200A191) 0 Revision 0                                                                      Page 63

Florida Power & Light             Attachments Florida Power & Light             Affachments ATTACHMENTS Page 64 Revision 00 Revision                               Page 64

S                                                           0                                                               E Florida Power & Light    Attachment A - NEI 04-02 Table B-1 Transition of Fundament Fire Protection Program & Design Elements A.      NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements 48 Pages Attached Page A-I Revision 0 0                                                                                                           Page A-1

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                       Statement  Compliance Basis 3.1 General                      3.1* General.                                                                 N/A        N/A - General statement, no technical requirements.
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.
3.2 Fire Protection Plan         N/A                                                                           N/A        Section Title 3.2.1 Intent                     3.2.1 Intent.                                                                 Complies 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 References                       Document ID O-ADM-016 Rev. 5 - Fire Protection Program 3.2.2 Management Policy          3.2.2* Management Policy Direction and Responsibility.                        Complies 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.
References                       Document ID O-ADM-016 Rev. 5 - Fire Protection Program 3.2.2.1 [Management Policy on   3.2.2.1*                                                                       Complies Senior Management]              The policy document shall designate the senior management position with immediate authority and responsibility for the fire protection program.
References                      Document ID O-ADM-016 Rev. 5 - Fire Protection Program UFSAR, Appendix 9.6A [Section 2.4] - Fire Protection Program Report 3.2.2.2 [Management Policy on   3.2.2.2*                                                                       Complies Daily Administration]           The policy document shall designate a position responsible for the daily administration and coordination of the fire protection program and its implementation.
References                      Document ID O-ADM-016 Rev. 5 [Section 3.8] - Fire Protection Program Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                            Run: 06/26/2012 11:05 Page: 1 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                        Statement  Compliance Basis 3.2.2.3 [Management Policy on    3.2.2.3*                                                                       Complies Interfaces]                      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.
References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 3.2.2.4 [Management Policy on    3.2.2.4*                                                                       Complies AHJ]                             The policy document shall identify the appropriate AHJ for the various areas of the fire protection program.
References                       Document ID 0-ADM-016 Rev. 5 [Section 3.8.7]- Fire Protection Program 3.2.3 Procedures                 3.2.3* Procedures.                                                             Complies 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:
References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 3.2.3 Procedures (1)             3.2.3 (1)
* Inspection, testing, and maintenance for fire protection systems   Complies   Procedures are established for inspection, testing and and features credited by the fire protection program                                      maintenance of fire protection systems as identified in Section 5.6 of 0-ADM-016.
Surveillance frequencies are outlined in Section 5.6 of 0-ADM-016 and may be modified in accordance with the methodology in EPRI Report TR1 006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide.
See Implementation Item 1 in Table S-3 of Attachment S.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.6] - Fire Protection Program Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                              Run: 06/26/2012 11:05       Page: 2 of 48 0                                                                                             0 0

0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                       Statement            Compliance Basis Open Item ID                     3.2.3(1)
Description                      Update the appropriate fire protection program document to provide a requirement that ifthe plant decides to implement the methodologies in EPRI Report TR1006756, that the methodology will be implemented in its entirety as it pertains to the element being implemented.
Date Entered Disposition Open                             Yes Corrective Action Include in LARITR                Yes 3.2.3 Procedures (2)             3.2.3 (2)
* Compensatory actions implemented when fire protection             Complies systems and other systems credited by the fire protection program and this standard cannot perform their intended function and limits on impairment duration References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 3.2.3 Procedures (3)             3.2.3 (3)
* Reviews of fire protection program - related performance and      Complies             Implementation Item - Plant documentation will be updated to trends                                                                                             add responsibilities for performing periodic program health monitoring. This will be done prior to the implementation date.
See Implementation Item 2 in Table S-3 of Attachment S.
References                       Document ID 0-ADM-501 Rev. 2 - Duties and Responsibilities of System Engineers 2006-1 App R - Appendix R Safe Shutdown Health Report 2006-2 System 016 Fire Protection Rev. 0 - Fire Protection System Health Report EDI-SE-005 - System Health Reports and Walkdowns NAP-407 Rev. 6 - Equipment Reliability Open Item ID                     3.2.3(3)
Description                      The monitoring program required by NFPA 805 will include a process that monitors and trends the fire protection program based on specific goals established to measure effectiveness.
Date Entered Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                          Run: 06/26/2012 11:05      Page: 3 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements / Guidance                                                       Statement  Compliance Basis Disposition Open                            Yes Corrective Action Include in LAR/TR               Yes 3.2.3 Procedures (4)            3.2.3 (4) Reviews of physical plant modifications and procedure changes       Complies for impact on the fire protection program References                      Document ID 0-ADM-104 Rev. 4 [Section 4.4.2.D] - 10 CFR 50.59 Applicability / Screening Reviews 0-ONOP-016.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions EN-AA-205-1100 Rev. 2 - Design Change Packages ENG-QI-01.7 Rev. 12 [Sections 5.2, 5.3] - Design Input/Verification ENG-QI-01.8 Rev. 28 [Section 5.1] - Design/Operability Reference Guide 3.2.3 Procedures (5)            3.2.3 (5) Long-term maintenance and configuration of the fire protection      Complies program References                      Document ID 0-ADM-016 Rev. 5 [Section 3] - Fire Protection Program 3.2.3 Procedures (6)            3.2.3 (6) Emergency response procedures for the plant industrial fire         Complies brigade.
References                      Document ID 0-ADM-016.2 - Fire Brigade Program 0-ONOP-016.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions 3.3 Prevention                  3.3 Prevention.                                                               Complies   The PTN fire prevention program is established and implemented A fire prevention program with the goal of preventing a fire from starting               as detailed in 0-ADM-016. It includes controls on operational shall be established, documented, and implemented as part of the fire                    activities and design controls that restrict the use of combustible protection program. The two basic components of the fire prevention                      materials. See following subsections for additional specific program shall consist of both of the following:                                          compliance statements and references.
(1) Prevention of fires and fire spread by controls on operational activities (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.
References                      Document ID 0-ADM-016 Rev. 5 - Fire Protection Program Fire S 6    nalysis Data Manager (4.1)                                                          Turkey Point                                                Run: 06/26/2012 11:05       Page: 4 of 48 0                                                                                  0

0                                                                                           0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                        Statement  Compliance Basis 3.3.1 Fire Prevention for        3.3.1 Fire Prevention for Operational Activities.                              Complies   See following subsections for additional compliance bases.
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.
References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 3.3.1.1 General Fire Prevention  3.3.1.1 General Fire Prevention Activities.                                    Complies   Procedures for the fire prevention have been developed and Activities                      The fire prevention activities shall include but not be limited to the                     implemented as part of the 0-ADM-016, Fire Protection Program.
following program elements:                                                               These procedures include reference to guidance regarding, but is not limited to the following programmatic elements of NFPA 805, section 3.3.1.1.
References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 3.3.1.1 General Fire Prevention  3.3.1.1 (1) Training on fire safety information for all employees and          Complies Activities (1)                   contractors including, as a minimum, familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms References                       Document ID 0-ADM-016 Rev. 5 [Section 3.7] - Fire Protection Program 3.3.1.1 General Fire Prevention 3.3.1.1 (2)
* Documented plant inspections including provisions for             Complies Activities (2)                  corrective actions for conditions where unanalyzed fire hazards are identified References                       Document ID 0-ADM-016.1 Rev. 1 [Section 3] -Transient Combustible and Flammable Substances Program QI-2-PTN-4 [Section 3] - Housekeeping 3.3.1.1 General Fire Prevention 3.3.1.1 (3)
* Administrative controls addressing the review of plant            Complies Activities (3)                  modifications and maintenance to ensure that both fire hazards and the impact on plant fire protection systems and features are minimized.
References                      Document ID 0-ADM-016 Rev. 5 - Fire Protection Program Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                               Run: 06/26/2012 11:05   Page: 5 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                       Statement   Compliance Basis 3.3.1.2 Control of Combustible   3.3.1.2* Control of Combustible Materials.                                    Complies    Procedures for the control of general housekeeping practices and Materials                        Procedures for the control of general housekeeping practices and the                      the control of transient combustibles have been developed and control of transient combustibles shall be developed and implemented.                     implemented as part of the O-ADM-016, Fire Protection Program These procedures shall include but not be limited to the following program                and administrative controls contained in 0-ADM-016.1. These elements:                                                                                 procedures include references to guidance regarding, but is not limited to the following programmatic elements of NFPA 805, section 3.3.1.2.
References                       Document ID C-ADM-016 Rev. 5 - Fire Protection Program 0-ADM-01 6.1 Rev. 1 - Transient Combustible and Flammable Substances Program 3.3.1.2 Control of Combustible   3.3.1.2 (1)
* Wood used within the power block shall be listed pressure-      Complies Materials (1)                    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.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1]- Fire Protection Program QI-2-PTN-4 [Section 5.6.8] - Housekeeping 3.3.1.2 Control of Combustible   3.3.1.2 (2) Plastic sheeting materials used in the power block shall be fire- Complies Materials (2)                    retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale tests, or equivalent.
References                       Document ID 0-ADM-016 Rev. 5 [5.1.1.10] - Fire Protection Program 3.3.1.2 Control of Combustible   3.3.1.2 (3) Waste, debris, scrap, packing materials, or other combustibles    Complies Materials (3)                    shall be removed from an area immediately following the completion of work or at the end of the shift, whichever comes first.
References                       Document ID 0-ADM-016 Rev. 5 - Fire Protection Program QI-2-PTN Housekeeping 3.3.1.2 Control of Combustible   3.3.1.2 (4)
* Combustible storage or staging areas shall be designated,       Complies Materials (4)                    and limits shall be established on the types and quantities of stored materials.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1.1.3] - Fire Protection Program 0-ADM-016.1 Rev. 1 [Section 4.3] - Transient Combustible and Flammable Substances Program Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                Run: 06/2612012 11:05    Page: 6 of 48 0                                                                                            0                                                                                0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference          Requirements I Guidance                                                    Statement     Compliance Basis 3.3.1.2 Control of Combustible    3.3.1.2 (5)
* Controls on use and storage of flammable and combustible     Complies via  Refer to the NFPA 30 Code Compliance Evaluation for specific Materials (5)                     liquids shall be in accordance with NFPA 30, Flammable and Combustible     Engineering   measures used for controlling combustible materials.
Liquids Code, or other applicable NFPA standards.                          Evaluation References                        Document ID 0-ADM-016 Rev. 5 - Fire Protection Program PTN-FPER-07-030 Rev. 1 - Code Compliance Evaluation, NFPA 30 - Flammable and Combustible Liquids Code Complies with Per FAQ 06-0020, the following guidance applies as to which Clarification 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."
No other NFPA standards were determined to be applicable.
References                        Document ID 0-ADM-016.1 Rev. 1 - Transient Combustible and Flammable Substances Program 3.3.1.2 Control of Combustible    3.3.1.2 (6)
* Controls on use and storage of flammable gases shall be in   Complies with PTN is not committed to any flammable gas standards, and as Materials (6)                     accordance with applicable NFPA standards.                                 Clarification such are not part of the current license 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."
Flammable gases are controlled per 0-ADM-016 and 0-ADM-016.1.
References                        Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 0-ADM-016.1 Rev. 1 - Transient Combustible and Flammable Substances Program 3.3.1.3 Control of Ignition      3.3.1.3 Control of Ignition Sources                                         N/A           Section Title Sources 3.3.1.3.1 [Control of Ignition   3.3.1.3.1"                                                                  Complies via  See code compliance evaluation for NFPA 51B.
Sources Code Requirements]       A hot work safety procedure shall be developed, implemented, and            Engineering periodically updated as necessary in accordance with NFPA 51 B,             Evaluation Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                  Run: 06/26/2012 11:05     Page: 7 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                          Statement     Compliance Basis References                       Document ID 0-ADM-016.5 Rev. 2 - Hot Work Program PTN-FPER-07-051 B Rev. 3 - Code Compliance Evaluation, NFPA 51 B - Fire Prevention in Use of Cutting and Welding Processes Complies with Compliance with NFPA 241 is addressed through compliance Clarification with NFPA 51 B. NFPA 241, 2000 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 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work."
3.3.1.3.2 [Control of Ignition   3.3.1.3.2                                                                        Complies      0-ADM-016 provides the following requirement:
Sources on Smoking               Smoking and other possible sources of ignition shall be restricted to Limitations]                     properly designated and supervised safe areas of the plant.                                    5.1.2 Control of Ignition Sources
: 1. Smoking is prohibited in the RCA, areas containing No Smoking signs, or any area containing easily ignitable material.
Smoking is allowed only in Designated Smoking Areas.
Other possible sources are addressed in 3.3.1.3 References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1.2] - Fire Protection Program 3.3.1.3.3 [Control of Ignition   3.3.1.3.3                                                                        Complies Sources for Leak Testing]        Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1.2] - Fire Protection Program 3.3.1.3.4 [Control of Ignition   3.3.1.3.4*                                                                       Complies Sources on Portable Heaters]     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.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1.2] - Fire Protection Program 3.3.2 Structural.                3.3.2 Structural.                                                                Complies 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.
References                       Document ID Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                 Run: 06/26/2012 11:05       Page: 8 of 48 0                                                                                               0                                                                                    40

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                          Statement          Compliance Basis STD-M-006 Rev. 1 - Turkey Point Unit 3 & 4 Fire Protection Standard 3.3.3 Interior Finishes          3.3.3 Interior Finishes.                                                         Complies via       The Control Room carpet has been tested to NFPA 253 (Critical Interior wall or ceiling finish classification shall be in accordance with       Engineering        Radiant Flux) and ASTM E-662 (Smoke Development) and is ML NFPA 101, Life Safety Code, requirements for Class A materials.                Evaluation         -Approved. PTN-FPER-96-001 indicates that the permantly Interior floor finishes shall be in accordance with NFPA 101 requirements                           installed combustible piping insulation has been evaluated to be for Class I interior floor finishes.                                                                an acceptable fire resistant material based on characteristics such as flame spread, ignition temperature, etc.
References                       Document ID 2006-29106-CR - Condition Report (Eval Complete) - Interior Finishes PTN-FPER-96-001 Rev. 0 - Evaluation of AP Armaflex Insulation for Condensation during Temporary Containment Cooling Complies           Insulation meets the applicable interior finish requirements.
Item for Implementation:
The applicable coating specifications will be updated as required to include the interior finish requirements.
See Implementation Item 3 in Table S-3 of Attachment S.
References                       Document ID 5177-P-161 Rev. 2 [Section 7.8] - Thermal Insulation for Piping and Equipment Outside Containment 5177-P-162 Rev. 3 [Section 7.7.2] - Technical Specification for Fabrication and Installation of Thermal Insulation Inside Containment License            In the OE review for Information Notice 2007-26 conducted under Amendment          CR 2007-26298, it was determined that certain self leveling floor Required           coatings have a flame spread rating of 30 when applied at a thickness of 49 mils. The evaluation also found that the thickness was installed at a thickness of up to 62 mils. This does not meet the requirement from NFPA 101-2006 that requires a flame spread rating of 25 or less.
See Attachment Lto the License Amendment Request.
References                       Document ID CR 2007-26298 - OE Review for Information Notice 2007-26 Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                         Run: 06/26/2012 11:05    Page: 9 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements I Guidance                                                       Statement               Compliance Basis Open Item ID                    3.3.3-1 Description                     Update the applicable coatings specifications to include the required combustibility ratings for interior finish from NFPA 101.
Date Entered Disposition Open                            Yes Corrective Action Include in LAR/TR               Yes 3.3.4 Insulation Materials      3.3.4 Insulation Materials.                                                   Complies via            See Reference Documents for Engineering Evaluations that Thermal insulation materials, radiation shielding materials, ventilation duct Engineering             demonstrate compliance with this requirement.
materials, and soundproofing materials shall be noncombustible or limited     Evaluation combustible.
References                      Document ID PTN-FPER-01-007 Rev. 0 - Evaluation of Combustible Loading due to Pipe Insulation PTN-FPER-96-001 Rev. 0 - Evaluation of AP Armaflex Insulation for Condensation during Temporary Containment Cooling Complies                5177-P-161 section 7.8 states the following: All materials shall be incombustible (sic). All insulation shall have a rating of 25 or less for flame spread, fuel consumption, and smoke generated when tested in accordance with ASTM-E-84.
5177-P-162, section 7.7.2 states the following: Thermal insulation materials shall be noncombustible or shall be tested to exhibit a rating of 25 or less for flame spread, fuel consumption, and smoke generated when tested in accordance with ASTM-E-84.
References                      Document ID 5177-P-161 Rev. 2 [Sec. 7.8] -Thermal Insulation for Piping and Equipment Outside Containment 5177-P-162 Rev. 3 [Sec. 7.7.2] - Technical Specification for Fabrication and Installation of Thermal Insulation Inside Containment 3.3.5 Electrical.               N/A                                                                           N/A                     N/A Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                            Run: 06/26/2012 11:05        Page: 10 of 48 0                                                                                           0                                                                                                 40

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                         Statement  Compliance Basis 3.3.5.1 [Electrical Wiring Above 3.3.5.1                                                                         License    Wiring above the control room ce iling does not meet this Suspended Ceiling Limitations]   Wiring above suspended ceiling shall be kept to a minimum. Where                Amendment  requirement.
installed, electrical wiring shall be listed for plenum use, routed in armored  Required cable, routed in metallic conduit, or routed in cable trays with solid metal               See Attachment Lto the License Amendment Request.
top and bottom covers.
Item for Implementation: Approp riate plant documentation will be revised to incorporate this require*ment for all future installations.
See Implementation Item 4 in Talble S-3 of Attachment S.
Open Item ID                     3.3.5.1 Description                      Update applicable cable specifications to reflect the requirements from 3.3.5.1.
Date Entered Disposition Open                             Yes Corrective Action Include in LAR/TR                Yes 3.3.5.2 [Electrical Raceway      3.3.5.2                                                                         License    Underground duct banks are PVC conduit encased in concrete.
Construction Limits]             Only metal tray and metal conduits shall be used for electrical raceways.       Amendment Thin wall metallic tubing shall not be used for power, instrumentation, or      Required   See Attachment L to the License Amendment Request.
control cables. Flexible metallic conduits shall only be used in short lengths to connect components.
References                       Document ID 5177-E-302 Rev. 10 [Sh. 3] - ELEC. INSTALLATION-RACEWAYS NOTES, SYS & DETAILS Complies   Exposed raceways not installed in underground duct banks at PTN comply with the requirements of this section.
References                       Document ID 5177-E-000 Rev. 3 - Electrical Design Criteria 5177-E-302 Sh 4 Rev. 5 - Electical Installation Raceways Notes, Symbols and Details 5177-E-302 Sh 5 Rev. 3 - Electrical Installation Raceways Notes, Symbols and Details Fire Safety Analysis Data Manager (4.1)                                                             Turkey Point                                              Run: 06/26/2012 11:05         Page: 11 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements / Guidance                                                        Statement            Compliance Basis 3.3.5.3 [Electrical Cable Flame 3.3.5.3*                                                                       Complies via         From Fire Protection SER dated 3/21/79 Propagation Limits]             Electric cable construction shall comply with a flame propagation test as      Previous acceptable to the AHJ.                                                         Approval             Section 3.1.14 Modifications "3.1.14 Fire Retardant Cable Coatings" Cables in trays with safe shutdown related cables in the following areas will be covered with a flame retardant coating:
(1) Inverter room (5.3)
(2) Containment (5.10)
(3) Lower level of turbine area (5.12)
(4) Auxiliary feed pump area (5.12)
(5) In the vicinity of the condensate storage tank and transfer pumps (5.13) 4.8 Electrical Cables "The cable insulation used in the plant consists of butyl rubber and polyethylene insulated conductors.
Cables routed in the containment have cross linked polyethylene insulated conductors. Most cable jackets are polyvinyl chloride with some cables having neoprene jackets. The flame test standard for cables, IEEE Std. 383-1974, was not in effect at the time cables were purchased and installed. Flame retardant coatings have been used extensively and in all areas of high cable concentration to reduce the fire hazards associated with combustible insulation materials. We find that the use of fire retardant coatings adequately addresses the combustible characteristics of cable insulation materials in some areas. With the exception of the open items (3.2.4, 3.2.5) this satisfies the objectives identified in Section 2.2 of this report and is therefore acceptable. We will report on these open items in a supplement to this report."
UFSAR Appendix 9.6A states "Plant design minimizes the use of combustible materials. Cables within certain areas are generally coated with a fire retardant coating or are qualified to IEEE 383-1974, or UL-910 per Reference 46" (PTN-FPER-05-001)
PTN maintains the coated cables and required new cables to be IEEE-383 or equivalent, therefore the bases for this approval have not been changed.
References                      Document ID Fire Protection Safety Evaluation Report Rev. 0 [Section 3.1.1.4] - Fire Protection Safety Evaluation Report Complies via         Refer to FPER-05-001 for documentation of acceptable flame Engineering          resistance standards for future cable installations.
Evaluation References                      Document ID 0
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                           Run: 06/26/2012 11:05 0Page: 12 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                        Statement       Compliance Basis PTN-FPER-05-001 Rev. 0 - Comparison of IEEE-383 and UL 910 Performance 3.3.6 Roofs.                     3.3.6 Roofs.                                                                   Complies        As stated in the UFSAR Section 9.6A, "Metal deck roof Metal roof deck construction shall be designed and installed so the roofing                    construction is noncombustible."
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                    The condensate polisher building has a steel roof with an be Class A as determined by tests described in NFPA 256, Standard                              insulated panel beneath it (drawing 5610-A-77), and field Methods of Fire Tests of Roof Coverings.                                                       walkdowns confirmed the C-Bus switchgear enclosures are metal walls and roof deck construction.
References                       Document ID 5610-A-77 Rev. 1 - Condensate Polishing Equipment Enclosure UFSAR, Appendix 9.6A -Fire Protection Program Report 3.3.7 Bulk Flammable Gas        3.3.7 Bulk Flammable Gas Storage.                                               Complies Storage.                         Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety.
References                       Document ID 0-ADM-016 Rev. 5 [Sec. 5.1.1.3.a] - Fire Protection Program 3.3.7.1 [Bulk Flammable Gas     3.3.7.1                                                                         Complies via    See NFPA 50A evaluation. Flammable gases are controlled per 0 Location Requirements]          Storage of flammable gas shall be located outdoors, or in separate              Engineering     -ADM-016 and 0-ADM-016.1.
detached buildings, so that a fire or explosion will not adversely impact       Evaluation systems, equipment, or components important to nuclear safety. NFPA                             Item for Implementation: Discrepancies identified in PTN-FPER-50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall                             07-050A will be dispositioned prior to program implementation.
be followed for hydrogen storage.                                                               See Implementation Item 5 in Table S-3 of Attachment S.
References                      Document ID 0-ADM-016 Rev. 5 - Fire Protection Program 0-ADM-016.1 Rev. 1 - Transient Combustible and Flammable Substances Program PTN-FPER-07-050A Rev. 0 - Code Compliance Evaluation NFPA 50A - Gaseous Hydrogen Systems Open Item ID                    3.3.7.1-1 Description                     Discrepancies identified in PTN-FPER-07-050A will be dispositioned prior to program implementation.
Date Entered Disposition Fire Safety Analysis Data Manager (4.1)                                                           Turkey P-oint                                                 Run: 06/26/2012 11:05      Page: 13 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                       Statement    Compliance Basis Open                             Yes Corrective Action Include in LARITR                Yes 3.3.7.2 [Bulk Flammable Gas      3.3.7.2                                                                       Complies Container Restrictions]          Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is not pointed at buildings.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.1.1.3] - Fire Protection Program 3.3.7.3 [Bulk Flammable Gas      3.3.7.3                                                                       Complies Cylinder Limitations]             Flammable gas storage cylinders not required for normal operation shall be isolated from the system.
References                       Document ID 0-ADM-016.1 Rev. 1 [Section 4.4.13] -Transient Combustible and Flammable Substances Program 3.3.8 Bulk Storage of            3.3.8 Bulk Storage of Flammable and Combustible Liquids.                      Complies     0-ADM-016.1 prohibits the bulk storage of combustible and Flammable and Combustible         Bulk storage of flammable and combustible liquids shall not be permitted                  flammable liquids in areas containing safety related Liquids.                         inside structures containing systems, equipment, or components important                   equipment. Storage of small amounts is allowed only if in a to nuclear safety. As a minimum, storage and use shall comply with NFPA                    combustible liquids storage cabinet.
30, Flammable and Combustible Liquids Code.
References                       Document ID 0-ADM-016.1 Rev. 1 [Section 5.3.4] -Transient Combustible and Flammable Substances Program PTN-FPER-07-030 Rev. 1 - Code Compliance Evaluation, NFPA 30 - Flammable and Combustible Liquids Code Complies via See NFPA 30 code compliance evaluation.
Engineering Evaluation References                        Document ID PTN-FPER-07-030 Rev. 1 - Code Compliance Evaluation, NFPA 30 - Flammable and Combustible Liquids Code License      See Attachment L to the License Amendment Request for Amendment    minor non-conformances with NFPA 30.
Required 3.3.9 Transformers.              3.3.9* Transformers.                                                          Complies 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.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                               Run: 06/26/2012 11:05     Page: 14 of 48 0                                                                                           0                                                                                 0

0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                        Statement           Compliance Basis References                       Document ID 0-ADM-034 Rev. 2 - Hazardous Material Emergency Response Plan and Oil Spill Prevention, Control and Countermeasure (SPCC) Plan 3.3.10 Hot Pipes and Surfaces. 3.3.10* Hot Pipes and Surfaces.                                                Complies 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.
References                       Document ID 0-ADM-01 6 Rev. 5 [Section 5.1.1.5] - Fire Protection Program 3.3.11 Electrical Equipment      3.3.11 Electrical Equipment                                                    Complies            Item for Implementation - Appropriate plant procedures will be Adequate clearance, free of combustible material, shall be maintained                              revised to provide instructions for providing adequate clearance around energized electrical equipment.                                                             free of combustible material around energized electrical equipment. This will be done prior to the implementation date.
See Implementation Item 6 in Table S-3 of Attachment S.
Open Item ID                     3.3.11 Description                      Provide guidance in a procedure to incorporate the resolution of FAQ 024. This states: "When used in Chapter 3, the term Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment is the clear space around equipment provided to ensure an acceptable level of fire prevention for Structures, Systems or Components (SSCs) necessary to ensure the Nuclear Safety Performance Criteria. This clear space or distance is maintained such that combustible material does not reside in an area where transient fuel packages have been shown to adversely affect energized electrical equipment needed to meet the nuclear safety performance criteria for the fire zone/area, either as an ignition source or target set."
Date Entered Disposition Open                             Yes Corrective Action Include in LAR/TR                Yes 3.3.12 Reactor Coolant Pumps     3.3.12* Reactor Coolant Pumps.                                                 Complies via       The ability of plant to withstand earthquake and other accident For facilities with non-inerted containments, reactor coolant pumps with an    Engineering        conditions is contained in FPER-11-014.
external lubrication system shall be provided with an oil collection system. Evaluation 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.
References                       Document ID PTN-FPER-11-014 Rev. 0 [Section 3.A, 3.B] - RCP Oil Collection System License Compliance Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                       Run: 06/26/2012 11:05      Page: 15 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference          Requirements / Guidance                                                         Statement    Compliance Basis Complies via The NRC concluded that oil collection system will withstand an Previous     SSE in letter dated 12/8/80:
Approval "In the SER, it was our concern that the reactor coolant pump oil collection system may not remain functional during an SSE.
By letter dated October 7, 1980, the licensee indicated that the oil collection system attachments will prevent the oil collection enclosures from falling and damaging safety related equipment. The licensee indicated that these enclosures will also remain functional during and SSE based on an analysis.
Based on the licensee's verification, we conclude that there is reasonable assurance that the oil collection system will withstand an SSE which meets Section 111.0 of Appendix R of 10 CFR Part 50 and, therefore, is acceptable" The oil collection system has not been modified and is consistent with the configuration approved in the SER.
References                        Document ID 1980-12 NRC-FPL Letter, Reactor Coolant Pump Oil Collection System 3.3.12 Reactor Coolant Pumps      3.3.12 (1) The oil collection system for each reactor coolant pump shall be     Complies via The ability of the system to collect oil from potential leakage (1)                               capable of collecting lubricating oil from all potential pressurized and        Engineering  points is documented in FPER-1 1-014.
nonpressurized leakage sites in each reactor coolant pump oil system.           Evaluation References                        Document ID PTN-FPER-11-014 Rev. 0 [Section 3.C] - RCP Oil Collection System License Compliance License      See Attachment Lto the License Amendment Request for oil Amendment    misting.
Required 3.3.12 Reactor Coolant Pumps     3.3.12 (2) Leakage shall be collected and drained to a vented closed             Complies via The ability of the system to collect the oil per this requirement is (2)                              container that can hold the inventory of the reactor coolant pump                Engineering  documented in FPER-11-014.
lubricating oil system.                                                         Evaluation References                        Document ID PTN-FPER-11-014 Rev. 0 [Section 3.E, 3.F] - RCP Oil Collection System License Compliance 3.3.12 Reactor Coolant Pumps     3.3.12 (3) A flame arrestor is required in the vent if the flash point           Complies via The justification for no flame arrestor is documented in FPER (3)                              characteristics of the oil present the hazard of a fire flashback.               Engineering  014.
Evaluation References                        Document ID PTN-FPER-11-014 Rev. 0 [Section 3.G] - RCP Oil Collection System License Compliance IFire Qa[VLY Analysis DaLa Manager k, II                                                               urmey o nL                                                 Run: 061261,20 12 1i:uu     rage: 16o148 0                                                                                              0                                                                                      0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                           Statement    Compliance Basis 3.3.12 Reactor Coolant Pumps     3.3.12 (4) Leakage points on a reactor coolant pump motor to be                   Complies via The ability of the collection system to collect from potential (4)                              protected shall include but not be limited to the lift pump and piping,           Engineering  leakage sites is documented in FPER-11-014.
overflow lines, oil cooler, oil fill and drain lines and plugs, flanged           Evaluation.
connections on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps.
References                       Document ID PTN-FPER-11-014 Rev. 0 [Section 3.C] - RCP Oil Collection System License Compliance 3.3.12 Reactor Coolant Pumps     3.3.12 (5) The collection basin drain line to the collection tank shall be        Complies via The ability of the drain line to accommodate the largest oil leak is (5)                              large enough to accommodate the largest potential oil leak such that oil          Engineering  documented in FPER-1 1-014.
leakage does not overflow the basin.                                              Evaluation References                       Document ID PTN-FPER-11-014 Rev. 0 [Section 3.D] - RCP Oil Collection System License Compliance 3.4 Industrial Fire Brigade. N/A                                                                               N/A          Section Title 3.4.1 On-Site Fire-Fighting      3.4.1 On-Site Fire-Fighting Capability.                                           N/A Capability                       All of the following requirements shall apply.
3.4.1 On-Site Fire-Fighting      3.4.1 (a) A fully staffed, trained, and equipped fire-fighting force shall be     Complies via The requirement states that a fire brigade is available "at all Capability (a)                   available at all times to control and extinguish all fires on site. This force    Previous     times". The PTN program allows the minimum complement to shall have a minimum complement of five persons on duty and shall                 Approval     drop for a maximum of two hours to account for unexpected conform with the following NFPA standards as applicable:                                       absences due to illness, family emergencies, etc.
The NRC approved the reduced fire brigade complement in an SER for Technical Specification revisions dated April 19, 1982. The approved Technical Specification for Fire Brigade states:
                                                                                                                                "A site Fire Brigade of at least 5 members shall be maintained onsite at all times*. The Fire Brigade shall not include 2 members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency."
                                                                                                                                "-Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accomodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements."
The fire brigade requirements have not been altered from the the approved technical specification requirements.
References                       Document ID Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                 Run: 06/26/2012 11:05       Page: 17 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference          Requirements I Guidance                                                    Statement          Compliance Basis 0-ADM-016 Rev. 5 [Section 5.6.6.1] - Fire Protection Program 0-ADM-016.2 [Section 4.2] - Fire Brigade Program 1982-04-19 [Page 6-5] - NRC-FPL Letter, Safety Evaluation for Amendment of Facility Operating Licenses including changes to the Fire Protection Technical Specifications Complies           A fully staffed, trained, and equipped fire brigade consisting of at least five members is maintained on site at all times (with the exception that composition may be less than the minimum requirements for a period of time not exceeding 2 hours). The Fire Brigade responds to all on site fires within the protected area and to a fire on site outside the protected area (i.e. nuclear training center, school/day care center, etc).
References                       Document ID 0-ADM-016 Rev. 5 [Section 3.16] - Fire Protection Program 0-ADM-01 6.2 [Section 4.2] - Fire Brigade Program 3.4.1 On-Site Fire-Fighting      3.4.1 (a)(1) NFPA 600, Standard on Industrial Fire Brigades (interior       Complies via       See NFPA 600 Code Compliance Evaluation Capability (a)(1)                structural fire fighting)                                                   Engineering Evaluation References                       Document ID PTN-FPER-07-600 Rev. 2 - Code Compliance Evaluation - NFPA 600, Standard on Industrial Fire Brigades 3.4.1 On-Site Fire-Fighting      3.4.1 (a)(2) NFPA 1500, Standard on Fire Department Occupational            N/A                NFPA 600 applies in lieu of NFPA 1500 Capability (a)(2)                Safety and Health Program References                       Document ID PTN-FPER-07-600 Rev. 2 - Code Compliance Evaluation - NFPA 600, Standard on Industrial Fire Brigades 3.4.1 On-Site Fire-Fighting      3.4.1 (a)(3) NFPA 1582, Standard on Medical Requirements for Fire           N/A                NFPA 600 applies in lieu of NFPA 1582 Capability (a)(3)                 Fighters and Information for Fire Department Physicians References                       Document ID PTN-FPER-07-600 Rev. 2 - Code Compliance Evaluation - NFPA 600, Standard on Industrial Fire Brigades 0
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                       Run: 06/26/2012 11:05        Page: 18 of 48 is                                                                                            is

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                       Statement       Compliance Basis 3.4.1 On-Site Fire-Fighting      3.4.1 (b)
* Industrial fire brigade members shall have no other assigned      Complies        The fire brigade shall not include the Shift Supervisor and the Capability (b)                   normal plant duties that would prevent immediate response to a fire or                        other members of the minimum shift crew necessary for safe other emergency as required.                                                                  shutdown of the unit and any personnel required for other essential functions during a fire emergency. The Fire Brigade Leader does not go into containment unless responding to a fire in containment. If no other non-Fire Brigade Operator is available, then a fire brigade member is permitted to go inside containment or to the switchyard.
References                       Document ID 0-ADM-016 Rev. 5 [Section 3.15, 5.6.6] - Fire Protection Program 0-ADM-016.2 [Section 3.5.6, 4.2.1] - Fire Brigade Program 3.4.1 On-Site Fire-Fighting      3.4.1 (c) During every shift, the brigade leader and at least two brigade     Complies Capability (c)                   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.
References                       Document ID 0-ADM-016.2 [Section 4.2.1]       - Fire Brigade Program 3.4.1 On-Site Fire-Fighting      3.4.1 (d) *The industrial fire brigade shall be notified immediately upon     Complies Capability (d)                   verification of a fire.
References                       Document ID O-EPIP-20101 Rev. 2 - Duties of Emergency Coordinator 0-ONOP-016.10 Rev. 8 [Section 4.0] - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions O-ONOP-016.8 Rev. 2 [Section 4.3, 5.1.2, 5.1.3] - Response to a Fire/Smoke Detection System Alarm 3.4.1 On-Site Fire-Fighting      3.4.1 (e) Each industrial fire brigade member shall pass an annual            Complies Capability (e)                   physical examination to determine that he 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.
References                       Document ID 0-ADM-01 6.2 [Section 5.3.1]     -  Fire Brigade Program Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                   Run: 06/26/2012 11:05      Page: 19 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                          Statement         Compliance Basis 3.4.2 Pre-Fire Plans.            3.4.2* Pre-Fire Plans.                                                           Complies 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.
References                       Document ID 0-ADM-016.2 - Fire Brigade Program 0-ONOP-016.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions 3.4.2.1 [Pre-Fire Plan Contents] 3.4.2.1*                                                                         Complies          Implementation Item -The pre-fire plans will be revised so that The plans shall detail the fire area configuration and fire hazards to be                          each pre-fire plan identifies the significant credible fire hazards encountered in the fire area, along with any nuclear safety components                             that may be encountered. In addition:
and fire protection systems and features that are present.
* Components necessary to achieve the nuclear safety performance criteria which require entry to the affected fire area will be included.
* The equipment and portions of the fire affected area where risk informed, performance-based analysis rely on assumptions that could be affected by fire brigade performance will be included.
This will be done prior to the implementation date.
See Implementation Item 7 in Table S-3 of Attachment S.
References                       Document ID 0-ONOP-01 6.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions Open Item ID                     3.4.2.1 Description                      Implementation Item - The pre-fire plans will be reviewed and updated accordingly to reflect the performance based program. This includes review/inclusion of the following:
* Components necessary to achieve the nuclear safety performance criteria which require entry to the affected fire area
* The equipment and portions of the fire affected area where risk-informed, performance-based analysis rely on assumptions that could be affected by fire brigade performance.
Date Entered Disposition Open                             Yes Corrective Action Run: 06/26/2012 11:05         Page: 20 of 48 is Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point 0                                                                                            0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                          Statement    Compliance Basis Include in LAR/TR                Yes 3.4.2.2 [Pre-Fire Plan Updates]  3.4.2.2                                                                          Complies Pre-fire plans shall be reviewed and updated as necessary.
References                       Document ID QI-5-PTN-1 Rev. 0 [Section 5.4.4] - Instructions, Procedures, and Drawings 3.4.2.3 [Pre-Fire Plan           3.4.2.3*                                                                         Complies Locations]                       Pre-fire plans shall be available in the control room and made available to the plant industrial fire brigade.
References                       Document ID QI-6-PTN-1 Rev. 4 [Section 2.4] - Document Control 3.4.2.4 [Pre-Fire Plan           3.4.2.4*                                                                         Complies Coordination Needs]              Pre-fire plans shall address coordination with other plant groups during fire emergencies.
References                       Document ID 0-ONOP-016.10 Rev. 8 [Section 5.3] - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions 3.4.3 Training and Drills        3.4.3 Training and Drills.                                                       Complies 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.
References                       Document ID 0-ADM-016.2 [Section 3.3.1] - Fire Brigade Program 3.4.3 Training and Drills (a)(1) 3.4.3 (a) Plant Industrial Fire Brigade Training. All of the following           Complies via See NFPA 600 Code Compliance Evaluation requirements shall apply.                                                        Engineering (1) Plant industrial fire brigade members shall receive training consistent      Evaluation with the requirements contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate.
References                       Document ID 0-ADM-016.2 - Fire Brigade Program PTN-FPER-07-600 Rev. 2 - Code Compliance Evaluation - NFPA 600, Standard on Industrial Fire Brigades Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                         Run: 06/26/2012 11:05 Page: 21 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                           Statement           Compliance Basis 3.4.3 Training and Drills (a)(2) 3.4.3 (a)(2) Industrial fire brigade members shall be given quarterly             Complies            Item for Implementation - The appropriate procedures will be training and practice in fire fighting, including radioactivity and health                            revised to contain a discussion about training in radioactivity and physics considerations, to ensure that each member is thoroughly familiar                             health physics considerations. This will be done prior to the with the steps to be taken in the event of a fire.                                                    implementation date.
See Implementation Item 8 in Table S-3 of Attachment S.
References                       Document ID O-ADM-016.2 - Fire Brigade Program Open Item ID                     3.4.3(a)(2)
Description                      Evaluate adding training in health physics and radiological hazards (specifically related to fire protection) to fire brigade training. Note: This may be combined with the brigade training updates for Radological Release.
Date Entered Disposition Open                             Yes Corrective Action Include in LARITR                Yes 3.4.3 Training and Drills (a)(3) 3.4.3 (a)(3) A written program shall detail the industrial fire brigade training  Complies program.
References                       Document ID O-ADM-016.2 - Fire Brigade Program 3.4.3 Training and Drills (a)(4) 3.4.3 (a)(4) Written records that include but are not limited to initial          Complies 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.
References                       Document ID O-ADM-O1 6.2 [Sec. 2.2] - Fire Brigade Program Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                        Run: 06/26/2012 11:05       Page: 22 of 48 0                                                                                               0                                                                                             0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                          Statement        Compliance Basis 3.4.3 Training and Drills (b)     3.4.3 (b) Training for Non-Industrial Fire Brigade Personnel. Plant              Complies         Implementation Item - The appropriate procedures will be personnel who respond with the industrial fire brigade shall be trained as                        updated to address the training of personnel who respond with to their responsibilities, potential hazards to be encountered, and                               the fire brigade. This will be done prior to the implementation interfacing with the industrial fire brigade,                                                     date.
See Implementation Item 9 in Table S-3 of Attachment S.
Open Item ID                      3.4.3(b)
Description                       Add requirements for training for security and other personnel required to show up at the scene of the fire to the appropriate program document..
Date Entered Disposition Open                              Yes Corrective Action Include in LAR/TR                 Yes 3.4.3 Training and Drills (c)(1)  3.4.3 (c)
* Drills. All of the following requirements shall apply.               Complies (1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire brigade.
References                        Document ID 0-ADM-016.2 [Sec. 5.3.1(1)] - Fire Brigade Program 3.4.3 Training and Drills (c)(2)  3.4.3 (c)(2) Industrial fire brigade drills shall be developed to test and       Complies 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.
References                        Document ID 0-ADM-016.2 [Sec. 5.4.8] - Fire Brigade Program 3.4.3 Training and Drills (c)(3)  3.4.3 (c)(3) Industrial fire brigade drills shall be conducted in various plant  Complies         Implementation Item - As part of the plant upgrade for areas, especially in those areas identified to be essential to plant                              compliance with NFPA 600, the appropriate procedures will be operation and to contain significant fire hazards.                                                revised to address this requirement. This will be done prior to the implementation date.
See Implementation Iteml0 in Table S-3 of Attachment S.
Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                     Run: 06/26/2012 11:05      Page: 23 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                          Statement       Compliance Basis Open Item ID                     3.4.3(c)(3)
Description                      As part of the plant upgrade for compliance with NFPA 600, the appropriate procedures will be revised to address this requirement.
Date Entered Disposition Open                             Yes Corrective Action Include in LARITR                Yes 3.4.3 Training and Drills (c)(4) 3.4.3 (c)(4) Drill records shall be maintained detailing the drill scenario,     Complies industrial fire brigade member response, and ability of the industrial fire brigade to perform as a team.
References                       Document ID 0-ADM-016.2 [Sec 5.4.9] - Fire Brigade Program 3.4.3 Training and Drills (c)(5) 3.4.3 (c)(5) A critique shall be held and documented after each drill.           Complies References                       Document ID 0-ADM-016.2 [Attachment 1] - Fire Brigade Program 3.4.4 Fire-Fighting Equipment. 3.4.4 Fire-Fighting Equipment.                                                   Complies        Implementation Item - The appropriate procedures will be revised Protective clothing, respiratory protective equipment, radiation monitoring                      to ensure more controls are in place to meet this equipment, personal dosimeters, and fire suppression equipment such as                           requirement. This will be done prior to the implementation date.
hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with                      See Implementation Item 11 in Table S-3 of Attachment S.
the applicable NFPA standards.
Open Item ID                     3.4.4.-l Description                      The appropriate procedures will be revised to ensure more controls are in place to ensure appropriate fire fighting equipment is provided to the fire brigade.
Date Entered Disposition Open                             Yes Corrective Action Include in LAR/TR                Yes Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                    Run: 06/26/2012 11:05       Page: 24 of 48 0                                                                                               0                                                                                        is

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference          Requirements / Guidance                                                            Statement  Compliance Basis 3.4.5 Off-Site Fire Department    N/A                                                                                N/A Interface.
3.4.5.1 Mutual Aid Agreement. 3.4.5.1 Mutual Aid Agreement.                                                      Complies Off-site fire authorities shall be offered a plan for their interface during fires and related emergencies on site.
References                        Document ID PTN-EP Rev. 45 - PTN Radiological Emergency Plan 3.4.5.2 Site-Specific Training. 3.4.5.2* Site-Specific Training.                                                    Complies 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.
References                        Document ID UFSAR, Appendix 9.6A [Section 3.9] - Fire Protection Program Report 3.4.5.3 Security and Radiation   3.4.5.3* Security and Radiation Protection.                                         Complies Protection.                      Plant security and radiation protection plans shall address off-site fire authority response.
References                       Document ID PTN-EP Rev. 45 - PTN Radiological Emergency Plan 3.4.6 Communications.            3.4.6* Communications.                                                              Complies An effective emergency communications capability shall be provided for the industrial fire brigade.
References                       Document ID 0-EPIP-20112 Rev. 0- Communication Network 0-ONOP-016.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe Shutdown Manual Actions 3.5 Water Supply                 N/A                                                                                 N/A        Section Title 3.5.1 [Water Supply Flow Code    3.5.1                                                                               Complies   PTN complies with Section (a).
Requirements]                    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 (1,1 35,500-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 Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                            Run: 06/26/2012 11:05 Page: 25 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements I Guidance                                                          Statement    Compliance Basis capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.
References                      Document ID 0-ADM-016 Rev. 5 [Sec. 5.3.1(a)] - Fire Protection Program 3.5.2 [Water Supply Tank Code   3.5.2*                                                                           Complies via Tanks are interconnected such that fire pumps can take suction Requirements]                   The tanks shall be interconnected such that fire pumps can take suction          Previous     from either or both. A failure in one tank or its piping would not from either or both. A failure in one tank or its piping shall not allow both    Approval     allow both tanks to drain.
tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.                                         Tanks are not designed in accordance with NFPA 22. NFPA 22 was not a committed code for this plant.
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                          The licensing basis for the water storage tanks is that contained lake), provided each fire pump has its own suction and both suctions and                      in Appendix A to BTP APCSB 9.5-1. This guideline stated: "Two pumps are adequately separated.                                                               separate, reliable water supplies should be provided. If tanks are used, two 100% (minimum of 300,000 gallons each) system Exception No. 2: Cooling tower basins shall be an acceptable water                            capacity tanks should be installed. They should be so interconnected that pumps can take suction from either or both.
source for fire pumps when the volume is sufficient for both purposes and However, a leak in one tank or its piping should not cause both water quality is consistent with the demands of the fire service.
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."
In addition, Section Ill.A of 10CFR50 Appendix R stated the following requirement. "Two separate water supplies shall be provided to furnish necessary water volume and pressure to the fire main loop.
Each supply shall consist of a storage tank, pump, piping, and appropriate isolation and control valves. Two separate redundant suctions in one or more intake structures from a large body of water (river, lake, etc.) will satisfy the requirement for two separated water storage tanks. These supplies shall be separated so that a failure of one supply will not result in a failure of the other supply.
Each supply of the fire water distribution system shall be capable of providing for a period of 2 hours the maximum expected water demands as determined by the fire hazards analysis for safety-related areas or other areas that present a fire exposure hazard to safety-related areas.
When storage tanks are used for combined service-water/fire-water uses the minimum volume for fire uses shall be ensured by means of dedicated tanks or by some physical means such as a vertical standpipe for other water service. Administrative controls, Fire Safety Analysis Data Manager (4.1)                                                             Turkey Point                                                   Run: 06/26/2012 11:05     Page: 26 of 48 0                                                                                              0                                                                                      0

0                                                                   0                                                                                   0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                              Statement  Compliance Basis including locks for tank outlet valves, are unacceptable as the only means to ensure minimum water volume.
Other water systems used as one of the two fire water supplies shall be permanently connected to the fire main system and shall be capable of automatic alignment to the fire main system.
Pumps, controls, and power supplies in these systems shall satisfy the requirements for the main fire pumps. The use of other water systems for fire protection shall not be incompatible with their functions required for safe plant shutdown. Failure of the other system shall not degrade the fire main system."
The fire water tanks are shared with the service water system. A vertical standpipe is provided as noted in the second paragraph from the BTP to provide the required capacity. Tanks shared between systems like this would not have been designed to NFPA requirements.
In the SER dated 3/21/1979, the NRC stated: "Fire water is supplied to the fire water pumps from a 500,000-gallon potable raw water storage tank which is filled by a connection to the municipal water system capable of 700 gpm makeup flow. A 100,000-gallon elevated storage tank is connected directly to the fire water mains. Since both of these water sources are used for other than fire water services, the quantity of water dedicated to fire water use is less that required to insure an adequate fire water supply. The licensee has agreed to administratively dedicate 180,000 gallons is water to fire protection purposes. We do not agree that administrative dedication of the fire water supply is an adequate measure of assuring that a proper fire water supply will be available. We have requested that an adequate supply of water be physically dedicated by vertical standpipe(s).
We find that, subject to implementation of the above described modification and resolution of the open items, the water supply may be adequate to meet the site fire water demand to protect safety-related systems."
In the SER dated 11/8/1981, the NRC provided the following statements Section III.A of Appendix R to 10 CFR Part 50 requires "two separate water supplies shall be provided to furnish necessary water volume and pressure to the main fire loop." In the case of Turkey Point this requires that FPL must, at a minimum (1) modify the existing on-site 500,000 gallon water tank by installing a new standpipe to dedicate a minimum of 300,000 gallons of the capacity of that tank to fire protection purposes; (2) design, engineer, procure, and construct a new on-site water tank of 750,000 gallon capacity, to include a redundant water supply of at least 300,000 gallons; Fire Safety Analysis Data Manager (4.1)                                  Turkey Point                                               Run: 06/26/2012 11:05       Page: 27 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements I Guidance                                                        Statement            Compliance Basis The new on-site water tank will be approximately 750,000 gallons capacity. It is necessary to install a tank of that size in order to both meet the NRC's requirements that a total of 600,000 gallons of water (300,000 gallons from each of the two on-site tanks) be dedicated for fire protection purposes from two separate, and redundant, sources and to supply the normal needs to the four generating units - two fossil and two nuclear - on the Turkey Point Site.
Based on our evaluation we conclude that FPL's request to be exempted from the schedular requirements of Section lILA "Water Supplies for Fire Suppression System" should be granted.
FPL's request for exemption from the scheduler requirements of Section III.A of Appendix R to 10 CFR Part 50 is hereby approved as described above."
The water tanks were installed in accordance with the details provided in the SER's and have not been modified, thereby maintaining the bases for the approval.
References                      Document ID 1981-11 NRC-FPL, Exemption to Certain Requirements of Section 50.48 and Appendix R to 10 CFR 50 561 0-M-3016, Sh. 3 Rev. 20 - Fire Protection System Units 1 thru 4, Fire Pumps Fire Protection Safety Evaluation Report Rev. 0 [Section 4.3.1.1] - Fire Protection Safety Evaluation Report UFSAR, Appendix 9.6A - Fire Protection Program Report 3.5.3 [Water Supply Pump        3.5.3*                                                                         Complies via         See NFPA 20 code compliance evaluation.
Code Requirements]              Fire pumps, designed and installed in accordance with NFPA 20,                 Engineering          See Reference Documents for additional Engineering Standard for the Installation of Stationary Pumps for Fire Protection, shall   Evaluation           Evaluations that demonstrate compliance with this requirement.
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.
References                      Document ID PTN-FPER-07-020 Rev. 0 - Code Compliance Evaluation, NFPA 20 - Standard for the Installation of Centrifugal Fire Pumps PTN-FPER-85-001 Rev. 0 - Fire Pump Controller Sensing Lines PTN-FPER-85-003 Rev. 0 - Diesel Engine Driven Fire Pump Heat Exchanger Piping PTN-FPER-85-005 Rev. 0 - Butterfly Valves Located in Fire Pump Suction Pipng 3.5.4 [Water Supply Pump        3.5.4                                                                          Complies             PTN uses one electric driven and one diesel driven pump.
Diversity and Redundancy]       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.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point 0
Run: 06/26/2012 11:05 is Page: 28 of 48

0                                                                               0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                      Statement  Compliance Basis References                       Document ID UFSAR, Appendix 9.6A [Section 2.4] - Fire Protection Program Report 3.5.5 [Water Supply Pump         3.5.5                                                                        License    The pumps are located in outdoor areas. The diesel driven fire Separation Requirements]         Each pump and its driver and controls shall be separated from the            Amendment  pump is located in a small nonrated enclosure. The electric remaining fire pumps and from the rest of the plant by rated fire barriers. Required   driven fire pump is exposed. The pumps are separated by a distance of greater than 20', with no exposed combustibles.
See Attachment L of the License Amendment Request.
3.5.6 [Water Supply Pump         3.5.6                                                                        Complies Startl/Stop Requirements]        Fire pumps shall be provided with automatic start and manual stop only.
References                       Document ID 5610-E-27 Sh 3Y1 Rev. 1 - Mechanical Auxiliaries Fire Pump Motor P39 Elementary Diagram J0259 Rev. 2 - Controller, Metron Inc.
3.5.7 [Water Supply Pump         3.5.7                                                                        Complies Connection Requirements]         Individual fire pump connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections.
References                       Document ID 5610-M-3016, Sh. 3 Rev. 20 - Fire Protection System Units 1 thru 4, Fire Pumps 3.5.8 [Water Supply Pressure     3.5.8                                                                        Complies Maintenance Limitations]        A method of automatic pressure maintenance of the fire protection water system shall be provided independent of the fire pumps.
References                       Document ID 5610-M-3016, Sh. 3 Rev. 20 - Fire Protection System Units 1 thru 4, Fire Pumps STD-M-006 Rev. 1 [Section 6.2.3.2] - Turkey Point Unit 3 & 4 Fire Protection Standard 3.5.9 [Water Supply Pump        3.5.9                                                                         Complies Operation Notification]          Means shall be provided to immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps.
References                       Document ID STD-M-006 Rev. 1 [Section 6.2.2] - Turkey Point Unit 3 & 4 Fire Protection Standard Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                              Run: 06/26/2012 11:05     Page: 29 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                        Statement      Compliance Basis 3.5.10 [Water Supply Yard        3.5.10                                                                         Complies via   See NFPA 24 Code Compliance Evaluation Main Code Requirements]          An underground yard fire main loop, designed and installed in accordance       Engineering with NFPA 24, Standard for the Installation of Private Fire Service Mains      Evaluation and Their Appurtenances, shall be installed to furnish anticipated water requirements.
References                       Document ID PTN-FPER-07-024 Rev. 1 - Code Compliance Evaluation - NFPA 24, Outside Protection PTN-FPER-10-003 Rev. 1 - Hose Testing Frequency PTN-FPER-93-006 Rev. 0 - Fire Protection Evaluation of NFPA-24 for Fire Main Under Dressout Building Extension 3.5.11 [Water Supply Yard        3.5.11                                                                         Compliance via The Fire Protection Safety Evaluation Report states: All yard Main Maintenance Issues]         Means shall be provided to isolate portions of the yard fire main loop for     Previous       hydrants, automatic and manual water systems are supplied by maintenance or repair without simultaneously shutting off the supply to        Approval       the fire loop. Sectionalyzing (sic) valves of the post-indicator type both fixed fire suppression systems and fire hose stations provided for                       are provided to isolate various sections of the loop. Hydrants are manual backup. Sprinkler systems and manual hose station standpipes                           all provided with sectional valves on the yard loop for isolation of shall be connected to the plant fire protection water main so that a single                   the hydrants in case of repair or maintenance. All post-indicator active failure or a crack to the water supply piping to these systems can be                  valves are administratively controlled by the use of seals and isolated so as not to impair both the primary and backup fire suppression                     periodic inspections made to verify that the valves are in the systems.                                                                                      proper position. Yard hydrants are strategically located around the exterior of the plant. Five hose houses are provided, each equipped with most of the necessary items for manual fire fighting. A few additional items will be provided. The hydrant hose threads are compatible with the local fire department. A 2 % -
inch gate valve will be provided for each hydrant. This will permit an additional hose to be connected to the hydrant without the need to shut down the hydrant... We find that, subject to implementation of the above described modifications, the fire water system is adequate to supply the fixed fire water suppression systems and manual hose stations. This satisfies the objectives identified in Section 2.2 of this report and is, therefore, acceptable.
Note, the modifications referred to in this approval include placing additional equipment at the hose houses. This does not impact the valves for isolation of the yard fire main loop. The yard fire main loop has not been modified which would invalidate the conclusions of the approval.
References                       Document ID Fire Protection Safety Evaluation Report Rev. 0 - Fire Protection Safety Evaluation Report Complies via   PTN-FPER-11-004 establishes the credited backup manual Engineering    suppression systems for those systems where an automatic Evaluation     suppression system shares the same supply with a manual suppression system. The credited systems are independent from the automatic suppression system.
0 Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                   Run: 06/26/2012 11:05        Page: 30 of 48 10                                                                                           0

0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                        Statement  Compliance Basis References                       Document ID PTN-FPER-1 1-004 Rev. 0 - Backup Suppression on Common Water Supply Systems 3.5.12 [Water Supply             3.5.12                                                                         Complies Compatible Thread                Threads compatible with those used by local fire departments shall be Connections]                     provided on all hydrants, hose 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.
References                       Document ID 0-ADM-016 Rev. 5 [Sections 5.3.1.3, 5.3.3.4.] - Fire Protection Program 3.5.13 [Water Supply Header      3.5.13                                                                         N/A        The plant does not use this arrangement.
Options]                         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.
3.5.14 [Water Supply Control     3.5.14*                                                                        Complies   PTN uses a combination of (a), (b), and (c).
Valve Supervision]               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.
(b) Locking valves in their normal position. Keys shall be made available only to authorized personnel.
(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.
References                       Document ID 0-OP-016.1 Rev. 2 [Section 1.2, Attachment 1] - Fire Protection Water System 561 0-M-3016 series - Fire Protection System drawings Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                            Run: 06/26/2012 11:05 Page: 31 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                       Statement    Compliance Basis 3.5.15 [Water Supply Hydrant     3.5.15                                                                        Complies Code Requirements]               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.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.3.3.3.4.a] - Fire Protection Program Complies via See NFPA 24 Code Compliance Evaluation for hose house Engineering  equipment compliance.
Evaluation References                       Document ID PTN-FPER-07-024 Rev. 1 - Code Compliance Evaluation - NFPA 24, Outside Protection 3.5.16 [Water Supply             3.5.16*                                                                       Complies     The plant complies with Exception No. 2.
Dedicated Limits]                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 protection demand for the specified duration as determined in this section.
References                       Document ID STD-M-006 Rev. 1 [Section 6.2.3.1] - Turkey Point Unit 3 & 4 Fire Protection Standard 3.6 Standpipe and Hose           N/A                                                                           N/A          Section Title Stations.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                              Run: 06/26/2012 11:05 Page: 32 of 48 0                                                                                            0                                                                             0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements / Guidance                                                    Statement      Compliance Basis 3.6.1 [Standpipe and Hose       3.6.1                                                                      Compliance via Class IIIhose stations are not provided. The original Appendix A Station Code Requirements]      For all power block buildings, Class IIIstandpipe and hose systems shall   Previous       requirement for hose stations are as follows:
be installed in accordance with NFPA 14, Standard for the Installation of  Approval Standpipe, Private Hydrant, and Hose Systems.                                             "Interior manual hose installation should be able to reach any location with at least one effective hose stream. To accomplish this, standpipes with hose connections equipped with a maximum of 75 feet of 1-1/2 inch woven jacket lined fire hose and suitable nozzles should be provided in all buildings, including containment, on all floors and should be spaced at not more than 100 foot intervals."
The UFSAR Section 9.6A Section 2.4 states: Manual hose stations with up to 100 feet of 1-1/2" minimum fire hose are located throughout the plant. The hoses are provided with matching fog nozzles. Hose stations are not generally located in radiological controlled areas or the control complex. Additional hose lengths may therefore be required for extremely remote areas.
3.

==2.3 DESCRIPTION==
Standpipes and Hose Stations Standpipe hose stations are provided in the plant (See Figures 9.6A-13 through 9.6A-1 5) so that any area containing essential equipment or components can be reached with an effective stream of water.
Two hose stations, HS-RW-01 and HS-RW-02, are connected to the service water system. These hose stations are equipped with a minimum of 75 feet of 1-1/2 inch (minimum) fire hose. Hose stations are not generally located in areas of radiological concern or the control complex. Thus, additional hose lengths may be required for remote areas. Most areas are within 30 feet of a fog nozzle. The RAB radioactive pipeway, the Containments, the Control Building, and the Diesel Generator Building may be reached with additional hose lengths. All nozzles are 1.5 inch (minimum), electrically safe fog nozzles. Threads on connections are those used by local fire departments."
The Fire Protection Safety Evaluation Report dated 3/21/79, states: There are 21 interior hose stations strategically located throughout the plant. Most hose stations are of the hose reel type equipped with 75 feet of 1-1/2 inch woven jacket lined fire hose.
All areas are accessible, within 20 feet of the fog nozzle, when hose lines are laid out. One-inch booster hose reels with low capacity nozzles and shutoff valves will be provided with sufficient hose reach for all areas of the control room, switchgear rooms and cable spreading room. We find that, subject to implementation of the above described modification, the interior hose stations are adequate to provide manual fire water suppression capability. This satisfies the objectives identified in Section 2.2 of this report and is therefore, acceptable."
Note: 1" booster hose stations were not provided. All areas of the control room, switchgear rooms and cable spreading rooms Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                   Run: 06/26/2012 11:05       Page: 33 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                         Statement    Compliance Basis are within reach of 1-1/2" hose stations, with adjustable flow nozzles that ensure low flow capabability. As a result, the intent of the modification is considered to be satisfied.
The standpipe system has not been modified and is consistent with the configuration approved in the SER.
References                       Document ID Fire Protection Safety Evaluation Report Rev. 0 - Fire Protection Safety Evaluation Report UFSAR, Appendix 9.6A [Section 2.4, 3.2.3] - Fire Protection Program Report Complies via See NFPA 14 Code Compliance Evaluation. Also, see Engineering  Reference Documents for Engineering Evaluations that Evaluation   demonstrate compliance with this requirement.
References                       Document ID PTN-FPER-07-014 Rev. 2 - Code Compliance Evaluation, NFPA 14 - Standpipe and Hose Stations PTN-FPER-85-002 Rev. 0 - Standpipe Hose Station Operating Pressure 3.6.2 [Standpipe and Hose        3.6.2                                                                           Complies via This was evaluated under PTN-FPER-85-002.
Station Capability Limitations]  A capability shall be provided to ensure an adequate water flow rate and        Engineering nozzle pressure for all hose stations. This capability includes the provision   Evaluation of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel.
References                       Document ID PTN-FPER-85-002 Rev. 0 - Standpipe Hose Station Operating Pressure 3.6.3 [Standpipe and Hose        3.6.3                                                                           Complies Station Nozzle Restrictions]     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.
References                       Document ID O-ADM-016 Rev. 5 [Section 5.3.3.4] - Fire Protection Program Fire Safety Analysis Data Manager (4.1) 0 Turkey Point 0
Run: 06/26/2012 11:05 is Page: 34 of 48

0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                       Statement    Compliance Basis 3.6.4 [Standpipe and Hose        3.6.4                                                                         Complies via Seismic standpipes are not an original commitment for this Station Earthquake Provisions]   Provisions shall be made to supply water at least to standpipes and hose      Previous     plant. See Section 3.6.1 for approval of the PTN hose stations.
stations for manual fire suppression in all areas containing systems and      Approval components needed to perform the nuclear safety functions in the event of                  The Federal Register notice that promulgated adoption of NFPA a safe shutdown earthquake (SSE).                                                          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 ABCSB 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 in accordance with Appendix A to BTP APCSB 9.5-1."
PTN received the construction permit prior to July 1, 1976 therefore this section is not applicable.
3.6.5 [Standpipe and Hose        3.6.5                                                                         N/A          Seismic hose stations are not provided.
Station Seismic Connection       Where the seismic required hose stations are cross-connected to Limitations]                     essential seismic non-fire protection water supply systems, the fire flow shall not degrade the essential water system requirement.
3.7 Fire Extinguishers.          3.7 Fire Extinguishers.                                                       Complies via See NFPA 10 code compliance evaluation.
Where provided, fire extinguishers of the appropriate number, size, and       Engineering type shall be provided in accordance with NFPA 10, Standard for Portable      Evaluation Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions.
References                       Document ID PTN-FPER-07-010 Rev. 1 - Code Compliance Evaluation - NFPA 10, Portable Fire Extinguishers 3.8 Fire Alarm and Detection     N/A                                                                           N/A          Section Title Systems.
3.8.1 Fire Alarm                 3.8.1 Fire Alarm.                                                             Complies via See NFPA 72 Series Code Compliance Evaluations Alarm initiating devices shall be installed in accordance with NFPA 72,       Engineering National Fire Alarm Code.Alarm annunciation shall allow the proprietary      Evaluation 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:
References                       Document ID Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                 Run: 06/26/2012 11:05     Page: 35 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference          Requirements I Guidance                                                        Statement            Compliance Basis PTN-FPER-07-072D Rev. 2 - Code Compliance Evaluation, NFPA 72D - Proprietary Protective Signaling Systems PTN-FPER-07-072E Rev. 0 - Code Compliance Evaluation, NFPA 72E - Automatic Fire Detectors 3.8.1 Fire Alarm (1)              3.8.1 (1) Actuation of any fire detection device                               Complies             Implementation Item -All areas in the power block will alarm in a constantly attended location after a modification to the fire detection system.
See Implementation Item 12 in Table S-3 of Attachment S.
References                        Document ID 0-ONOP-016.8 Rev. 2 - Response to a Fire/Smoke Detection System Alarm Open Item ID                      3.8.1(1)-i Description                       During the modification to the fire detection system, ensure all alarms annunciate in a constantly attended location.
Date Entered Disposition Open                              Yes Corrective Action Include In LAR/TR                 Yes 3.8.1 Fire Alarm (2)              3.8.1 (2) Actuation of any fixed fire suppression system                       License              Water flow alarms or trip horns/bells are provided for each water Amendment            suppression in the power block. However, not all of these fire Required             suppression systems in the nuclear power block alarm at a constantly attended location See Attachment Lof the License Amendment Request.
References                         Document ID 3-ARP-097.CR.G Rev. 0 - CONTROL ROOM RESPONSE - PANEL G 3-ARP-097.CR.I Rev. 0 - CONTROL ROOM RESPONSE - PANEL I 3.8.1 Fire Alarm (3)              3.8.1 (3) Actuation of any manual fire alarm station                           Complies References                         Document ID 0-ONOP-01 6.8 Rev. 2 - Response to a Fire/Smoke Detection System Alarm Fire a6       nalysis Data Manager (4.1)                                                          Turkey Point                                                        Run: 06/26/2012 11:05        Page: 36 of 48 0                                                                                             is

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                       Statement  Compliance Basis 3.8.1 Fire Alarm (4)             3.8.1 (4) Starting of any fire pump                                           Complies References                       Document ID O-ARP-097.WTP - Water Treatment Plant Annunciator Response 3.8.1 Fire Alarm (5)             3.8.1 (5) Actuation of any fire protection supervisory device                 Complies References                       Document ID 0-ARP-097.WTP - Water Treatment Plant Annunciator Response 3-ARP-097.CR.X Rev. 0 - CONTROL ROOM RESPONSE - PANEL X 3.8.1 Fire Alarm (6)             3.8.1 (6) Indication of alarm system trouble condition                        Complies References                       Document ID 0-ONOP-016.8 Rev. 2 - Response to a Fire/Smoke Detection System Alarm 3.8.1.1 [Fire Alarm              3.8.1.1                                                                       Complies Communication Requirements]      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.
References                       Document ID 0-EPIP-20112 Rev. 0- Communication Network 3.8.1.2 [Fire Alarm Prompt       3.8.1.2                                                                       N/A        General Statement - No Technical Requirements Notification Limits]             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:
3.8.1.2 [Fire Alarm Prompt       3.8.1.2 (1) General site population in all occupied areas                     Complies Notification Limits] (1)
References                       Document ID 0-EPIP-20112 Rev. 0 [Section 4.1] - Communication Network 3.8.1.2 [Fire Alarm Prompt       3.8.1.2 (2) Members of the industrial fire brigade and other groups           Complies Notification Limits] (2)         supporting fire emergency response References                       Document ID 0-EPIP-20112 Rev. 0 [Section 4.5.1]    - Communication Network Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                          Run: 06/26/2012 11:05 Page: 37 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                       Statement      Compliance Basis 3.8.1.2 [Fire Alarm Prompt       3.8.1.2 (3) Off-site fire emergency response agencies. Two independent        Complies Notification Limits] (3)         means shall be available (e.g., telephone and radio) for notification of off-site emergency services References                       Document ID 0-EPIP-20112 Rev. 0- Communication Network 3.8.2 Detection.                 3.8.2 Detection.                                                              Complies via   See NFPA 72 series code compliance evaluations. Also, see If automatic fire detection is required to meet the performance or            Engineering    Reference Documents for Engineering Evaluations that deterministic requirements of Chapter 4, then these devices shall be          Evaluation     demonstrate compliance with this requirement. See Table 4-3 of installed in accordance with NFPA 72, National Fire Alarm Code, and its                      the LAR for required systems.
applicable appendixes.
References                       Document ID PTN-FPER-01-008 Rev. 0 - Evaluation of Fire Detection Capability PTN-FPER-02-001 Rev. 0 - Evaluation of Fire Detection Capability with OOS Loop PTN-FPER-04-002 Rev. 0 - Alarm Point 3 Loop OOS PTN-FPER-07-002 Rev. 0 - Alternate Detection for CCW Equipment Areas PTN-FPER-07-072D Rev. 2 - Code Compliance Evaluation, NFPA 72D - Proprietary Protective Signaling Systems PTN-FPER-07-072E Rev. 0 - Code Compliance Evaluation, NFPA 72E - Automatic Fire Detectors PTN-FPER-10-008 Rev. 0 - EOL Device at Detector 01-16 PTN-FPER-11-008 Rev. 0 - Suppression and Detection in Outdoor Areas 3.9 Automatic and Manual         N/A                                                                           N/A            Section Title Water-Based Fire Suppression Systems.
3.9.1 [Fire Suppression System   3.9.1*                                                                        Complies via   See NFPA 13 and 15 code compliance evaluations. Also, see Code Requirements]               If an automatic or manual water-based fire suppression system is required     Engineering    Reference Documents for Engineering Evaluations that to meet the performance or deterministic requirements of Chapter 4, then      Evaluation     demonstrate compliance with this requirement.No water mist or the system shall be installed in accordance with the appropriate NFPA                        foam systems are installed. See Table 4-3 of the LAR for standards including the following:                                                           required systems.
(1) NFPA 13, Standard for the Installation of Sprinkler Systems (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection (3) NFPA 750, Standard on Water Mist Fire Protection Systems (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems References                       Document ID PTN-FPER-07-013 Rev. 0 - Code Compliance Evaluation, NFPA 13 - Standard for the Installation of Sprinkler Systems, 4A/4B EDG Building PTN-FPER-07-015 Rev. 0 - Code Compliance Evaluation, NFPA 15 - Standard for Water Spray Fixed Systems for Fire Protection PTN-FPER-11-008 Rev. 0 - Suppression and Detection in Outdoor Areas PTN-FPER-85-004 Rev. 0 - Fire Suppression System - North / South Breezeway Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                               Run: 06/26/2012 11:05     Page: 38 of 48 0                                                                                 0

0                                                                                           0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                        Statement    Compliance Basis PTN-FPER-97-015 Rev. 0 - Technical Evaluation for Turbine Building Fire Protection PTN-FPER-98-014 Rev. 0 - Evaluation of Charging Pump Room Partial Area Suppression 3.9.2 [Fire Suppression System   3.9.2                                                                          Complies Flow Alarm]                      Each system shall be equipped with a water flow alarm.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.3.2, 5.6.4.1.a(3)] - Fire Protection Program 3.9.3 [Fire Suppression System   3.9.3                                                                          License      Not all of the suppression systems in the power block provide a Alarm Locations]                All alarms from fire suppression systems shall annunciate in the control        Amendment    remote annunciation upon actuation. See Attachment L of the room or other suitable constantly attended location.                           Required     License Amendment Request.
3.9.4 [Fire Suppression System   3.9.4                                                                          License      See Attachment L of the License Amendment Request.
Diesel Pump Sprinkler            Diesel-driven fire pumps shall be protected by automatic sprinklers.           Amendment Protection]                                                                                                     Required 3.9.5 [Fire Suppression System   3.9.5                                                                          Complies Shutoff Controls]                Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve.
References                       Document ID 0-ADM-016 Rev. 5 [Section 5.3.2] - Fire Protection Program 3.9.6 [Fire Suppression System   3.9.6                                                                          Complies     See section 3.5.14 compliance basis.
Valve Supervision]              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.
3.10 Gaseous Fire Suppression    N/A                                                                            N/A          Section Title Systems.
3.10.1 [Gaseous Suppression      3.10.1                                                                         Complies via See NFPA 12A Code Compliance Evaluation.
System Code Requirements]        If an automatic total flooding and local application gaseous fire              Engineering suppression system is required to meet the performance or deterministic         Evaluation   There are no clean agent or carbon dioxide systems installed.
requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes:                           See Table 4-3 of the LAR for required systems.
(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems (2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems References                       Document ID PTN-FPER-07-012A Rev. 1 - Code Compliance Evaluation, NFPA 12A - Halon 1301 Fire Extinguishing Systems Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                              Run: 06/26/2012 11:05      Page: 39 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                       Statement    Compliance Basis 3.10.2 [Gaseous Suppression      3.10.2                                                                        Complies via See NFPA 72D/E evaluations.
System Alarm Location]           Operation of gaseous fire suppression systems shall annunciate and            Engineering alarm in the control room or other constantly attended location identified. Evaluation References                       Document ID PTN-FPER-07-072D Rev. 2 - Code Compliance Evaluation, NFPA 72D - Proprietary Protective Signaling Systems PTN-FPER-07-072E Rev. 0 - Code Compliance Evaluation, NFPA 72E - Automatic Fire Detectors 3.10.3 [Gaseous Suppression      3.10.3                                                                        Complies System Ventilation Limitations]  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.
References                       Document ID PC/M 83-139 Rev. 4 - Halon Suppression Systems for Appendix R Modifications 3.10.4 [Gaseous Suppression      3.10.4*                                                                       N/A          There are no backup gaseous system.
System Single Failure Limits]    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.
3.10.5 [Gaseous Suppression      3.10.5                                                                        Complies System Disarming Controls]       Provisions for locally disarming automatic gaseous suppression systems shall be secured and under strict administrative control.
References                       Document ID 0-ADM-016.3 Rev. 1 - Fire Protection Impairments 3.10.6 [Gaseous Suppression      3.10.6"                                                                       N/A          C02 systems are not used at the plant.
System C02 Limitations]          Total flooding carbon dioxide systems shall not be used in normally occupied areas.
3.10.7 [Gaseous Suppression      3.10.7                                                                        N/A          C02 systems are not used at the plant.
System C02 Warnings]            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.
3.10.8 [Gaseous Suppression      3.10.8                                                                        N/A          C02 systems are not used at the plant.
System C02 Required              Positive mechanical means shall be provided to lock out total flooding Disarming]                       carbon dioxide systems during work in the protected space.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                             Run: 06/26/2012 11:05 Page: 40 of 48 0
0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                        Statement       Compliance Basis 3.10.9 [Gaseous Suppression     3.10.9                                                                          Complies System Cooling                  The possibility of secondary thermal shock (cooling) damage shall be Considerations]                 considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide.
References                      Document ID PC/M 83-139 Rev. 4 - Halon Suppression Systems for Appendix R Modifications 3.10.10 [Gaseous Suppression    3.10.10                                                                         Complies via    See PTN-FPER-1 1-005 for this evaluation.
System Decomposition Issues]    Particular attention shall be given to corrosive characteristics of agent       Engineering decomposition products on safety systems.                                       Evaluation References                      Document ID PTN-FPER-1 1-005 Rev. 0 - Evaluation of the Corrosive Effects of Halon 1301 Decomposition Products 3.11 Passive Fire Protection    3.11 Passive Fire Protection Features.                                          N/A             N/A - General statement, no technical requirements.
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.
3.11.1 Building Separation. 3.11.1 Building Separation.                                                     Complies via    See the engineering evaluations listed in the reference Each major building within the power block shall be separated from the          Engineering     documents for performance based evaluations.
others by barriers having a designated fire resistance rating of 3 hours or     Evaluation 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.
References                      Document ID PTN-FPER-89-009 Rev. 0 - Justification for not Sealing the Exterior Wall/Roof Penetrations which Separate Redundant Safe Shutdown Equipment PTN-FPER-89-023 Rev. 3 - Earth Bound Walls Designated as Fire Barriers Complies        In general, buildings are separated by 3-hour fire resistance rated barriers or 50 feet of open space.
References                      Document ID UFSAR, Appendix 9.6A [Section 3.11.2] - Fire Protection Program Report Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                  Run: 06/26/2012 11:05       Page: 41 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                          Statement      Compliance Basis 3.11.2 Fire Barriers.            3.11.2 Fire Barriers.                                                            Complies       Walls, floors and ceilings separating Fire Areas shall be fire Fire barriers required by Chapter 4 shall include a specific fire-resistance                    barriers with a minimum rating of three hours as defined by rating. Fire barriers shall be designed and installed to meet the specific fire                 ASTM E-1 19 (e.g. six inch solid concrete or greater, 8 inch resistance rating using assemblies qualified by fire tests. The qualification                   concrete filled concrete block or greater, unless the barrier is fire tests shall be in accordance with NFPA 251, Standard Methods of                            subject to an engineering equivalency evaluation.
Tests of Fire Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials.
References                       Document ID UFSAR, Appendix 9.6A [3.11.2] - Fire Protection Program Report Complies via   See reference documents for engineering evaluations.
Engineering Evaluation References                       Document ID PTN-FPER-02-003 Rev. 0 - Fire Barrier Inspection of Containment Walls PTN-FPER-04-001 Rev. 0 - Electrical Manhole Cover Operability PTN-FPER-08-010 Rev. 0 - Fire Area DDD and EEE Boundaries PTN-FPER-08-011 Rev. 0 - Fire Area CCC Boundaries PTN-FPER-1 1-010 Rev. 0 - Acceptability of Unsealed Penetrations PTN-FPER-88-001 Rev. 1 - Fire Protection Evaluation for Inspection Criteria for Cracks in Masonry Fire Walls PTN-FPER-89-005 Rev. 1 - Fire Area Boundary Evaluation - Fire Barriers 72W & 73W PTN-FPER-89-015 Rev. 0 - Containment Electrical Penetrations PTN-FPER-90-007 Rev. 0 - Protection of Embedded Conduit in Unit 4 EDG Building 4D Switchgear Room PTN-FPER-90-008 Rev. 0 - Protection Of Embedded Conduit In Unit 4 EDG Room and Fuel Oil Transfer Pump Room PTN-FPER-91-001 Rev. 0 - UNIT 4 EDG Building Exterior Wall Penetrations PTN-FPER-96-023 Rev. 0 - Evaluation of the Need for Passive Fire Protection for Structural Steel in the Control Building Elevator Vestibule Area (Fire Zone 64/
Fire Area CCC) 3.11.3 Fire Barrier              3.11.3* Fire Barrier Penetrations.                                               Complies via   See Code Compliance Evaluations for NFPA 80 and 90A.
Penetrations.                    Penetrations in fire barriers shall be provided with listed fire-rated door      Engineering    Additional evaluations are shown in the Reference Documents.
assemblies or listed rated fire dampers having a fire resistance rating          Evaluation 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 (2) NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating Systems Fire Safety Analysis Data Manager (4.1) 0                                                                                              Turkey Point 0
Run: 06/26/2012 11:05 0
Page: 42 of 48

0                                                                                              0                                                                               0 Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements I Guidance                                                          Statement    Compliance Basis (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.
References                      Document ID PTN-FPER-00-002 Rev. 0 [Entire Document] - Evaluation of Fire Door D046-2 Modifications PTN-FPER-01-003 Rev. 0 - Evaluation of Electronic Monitoring of Fire Doors PTN-FPER-02-002 Rev. 0 - Evaluation of Control Room Door D1 06-2 Cracks PTN-FPER-07-001 Rev. 0 - Hot Lab HVAC Openings 022S1-H002 & 022S1 -H003 PTN-FPER-07-080 Rev. 1 - Code Compliance Evaluation, NFPA 80 - Fire Doors and Windows PTN-FPER-07-090A Rev. 1 - Code Compliance Evaluation, NFPA 90A - Installation of Air Conditioning and Ventilation Systems (Fire Dampers)
PTN-FPER-09-003 Rev. 0 - Spurious Disengagement of Security Door Latches PTN-FPER-10-001 Rev. 0 - D075-1 Frame Fire Resistance Restoration PTN-FPER-1 1-003 Rev. 0 - Turkey Point GL 86-10 Evaluation of Adequacy of Radiation Shield Doors in Charging Pump Rooms for Preventing Spread of Fire PTN-FPER-86-001 Rev. 0 - Control Room Door D1 06-1 PTN-FPER-87-001 Rev. 0 - Machine Shop Rollup Door D25-2 PTN-FPER-89-001 Rev. 4 - Fire Protection Evaluation for Minor Modifications and Attachments to Fire Doors PTN-FPER-89-002 Rev. 1 - Auxiliary Building Roof Penetration 28C-H002 PTN-FPER-89-003 Rev. 0 - Conduit Penetrations on the Door Frame for Door No. 63-2 PTN-FPER-89-006 Rev. 0 - Concealed Electrical Cable Fire Protection PTN-FPER-89-013 Rev. 1 - Concrete Hatch Equipment Covers PTN-FPER-89-017 Rev. 0 - Floor Mounted Reach Rod Assemblies PTN-FPER-89-018 Rev. 1 - Electrical Penetration Room Fire Dampers PTN-FPER-89-019 Rev. 1 - Containment Mechanical Penetrations PTN-FPER-90-005 Rev. 0 - Electrical Equipment Room Exterior Wall Penetrations PTN-FPER-92-004 Rev. 0 - Auxiliary Building Roof Hatch Evaluation PTN-FPER-93-002 Rev. 1 - Fire Protection Evaluation for Turbine Pedestal to Turbine Deck Expansion Joint PTN-FPER-97-020 Rev. 0 - Repealing the Fire Barrier Requirements Assigned to Penetration No. 041 E-EO01 in the East Wall of the Boric Acid Storage Tank Room Fire Safety Analysis Data Manager (4.1)                                                             Turkey Point                                             Run: 06/26/2012 11:05     Page: 43 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference        Requirements / Guidance                                                            Statement     Compliance Basis Complies with NFPA 101, Section 8.2.3.2.1(a) with regards to rated fire door Clarification assemblies refers to NFPA 80. NFPA 101 Section 9.2.1 with regards to rated fire dampers refers to NFPA 90A. PTN has performed detailed code compliance reviews for both NFPA 80 and NFPA 90A. NFPA 101 compliance is achieved through compliance with NFPA 80 and NFPA 90A.
3.11.4 Through Penetration      3.11.4* Through Penetration Fire Stops.                                            Complies      Section (a): Specification 5177-265-A-1 50 states the following Fire Stops.                     Through penetration fire stops for penetrations such as pipes, conduits,                         regarding qualification testing:
bus ducts, cables, wires, pneumatic tubes and ducts, and similar building                        2.3 Conformance to the following codes and standards shall be service equipment that pass through fire barriers shall be protected as follows.                                                                                         met by the Supplier:
(a) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive                      2.3.1 American Society for Testing and Materials (ASTM) penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a                      E 119 Fire Tests of Building Construction and Materials fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.                                            2.3.2 Nuclear Regulatory Commission (NRC) - Regulatory (b) Conduits shall be provided with an internal fire seal that has an                            Guides 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                   10CFR50 Appendix R Fire Protection Program for Nuclear Power location that is as close to the barrier as possible.                                            Plants, Section J, Fire Barrier Penetration Seal Qualification 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                   2.3.4 Institute of Electrical and Electronics Engineers (IEEE) conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In Standard 634 Standard Cable Penetration Fire Stop Qualification this case the conduit opening shall be provided with noncombustible Test 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.
References                      Document ID 5177-265-A-1 50 Rev. 2 -TECHNICALSPECIFICATION FOR DESIGN, TECHNICAL SERVICES, AND FURNISHING OF ELASTOMER PENETRATION SEALS AND BLOCKOUT CLOSURES FOR FIRE PROTECTION Complies via  FPER-96-024 provides the methodology to evaluate penetration Engineering   seals against the limiting parameters to establish the fire barrier Evaluation    rating. The critical parameters that are reviewed to ensure that a seal is bounded by a test area as follows: Barrier Construction I Thickness Maximum Opening Size and Type Seal Material and Depth Damming Material and Thickness Maximum Free Area Penetrant Spacing (Max. & Min.) Seal Orientation Maximum Pipe Temperature and Movement Cable Fill and Type Number, Size and Type of Penetrants References                      Document ID PTN-FPER-96-024 (Funct. Equiv.) Rev. 0 [Entire Document] - Technical evaluation to compare a sample population of penetration fire seals against tested configurations Fire S 6     nalysis Data Manager (4.1)                                                               Turkey Point                                                  Run: 06/26/2012 11:05 Is Page: 44 of 48 0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements / Guidance                                                  Statement    Compliance Basis Complies via Evaluations of penetration seal deviations are listed in the Engineering  Reference Documents.
Evaluation References                       Document ID PTN-FPER-00-001 Rev. 0 - Evaluation of Conduit Stuffing PTN-FPER-01-006 Rev. 0 - Evaluation of Inaccessible Penetration Seals PTN-FPER-1 1-010 Rev. 0 - Acceptability of Unsealed Penetrations PTN-FPER-86-002 Rev. 0 - Seismic Joint - Dymeric Seal Details PTN-FPER-89-010 Rev. 0 - Penetration No. 070C-E007; Non-Rated Seals PTN-FPER-89-011 Rev. 0 - Penetration No. 067C - E013; Non-Rated Seals PTN-FPER-89-014 Rev. 2 - Control Room Floor Penetrations PTN-FPER-89-016 Rev. 1 - Control Room CMU Wall Cable Tray Penetration Seals PTN-FPER-89-020A Rev. 0 - Penetration Seal 016W-M005 PTN-FPER-89-020B Rev. 0 - Penetration Seal 013S-MOO1 PTN-FPER-89-020C Rev. 1 - Penetration Seal Design Deviation PTN-FPER-89-020D Rev. 1 - Penetration Seal Design Deviation PTN-FPER-89-020F Rev. 0 - Penetration Seal Design Deviation 012S-M006 PTN-FPER-89-020G Rev. 1 - Penetration Seal Design Deviation PTN-FPER-89-020H Rev. 1 - Penetration Seal Design Deviation PTN-FPER-89-0201 Rev. 1 - Penetration Seal Design Deviation PTN-FPER-89-020J Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020K Rev. 2 - Reach Rod Penetrations Through Fire Barriers PTN-FPER-89-020L Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020M Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020N Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020P Rev. 0 - Electrical Penetrations Using Primer Coated Cable PTN-FPER-89-020Q Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020R Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020S Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020T Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020U Rev. 0 - Penetration Seal Design Deviation PTN-FPER-89-020V Rev. 0 - Penetration Seal Design Deviation PTN-FPER-90-002 Rev. 0 - Penetration Seal Design Deviation PTN-FPER-90-006 Rev. 0 - Unit 4 EDG Building Roof Penetrations Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                              Run: 06/26/2012 11:05        Page: 45 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                          Statement      Compliance Basis PTN-FPER-96-024 (Funct. Equiv.) Rev. 0 - Technical evaluation to compare a sample population of penetration fire seals against tested configurations PTN-FPER-96-034 Rev. 0 - Technical Evaluation for the Qualification of KBS Sealbags for use as a Temporary Through Wall Penetration Fire Stop PTN-FPER-97-014 Rev. 0 - Evaluation of Modified Fire Penetration Boot Seal for Penetration 038F-MOO8 PTN-FPER-98-015 Rev. 0 - Evaluation of Seals with Ceramic Fiber PTN-FPER-98-016 Rev. 0 - Evaluation of Fire Rated Penetration Seals with Large Pipes PTN-FPER-98-017 Rev. 0 - Evaluation of Fire Rated Penetration Seals with Structural Steel PTN-FPER-98-018 Rev. 0 - Evaluation of Fire Rated Penetration Seals with Electrical Penetrants PTN-FPER-98-019 Rev. 0 - Evaluation of Seals with Caulk, Foam, T-Lag or Flamemastic PTN-FPER-99-005 Rev. 13 - Penetration Seal Evaluations Complies via   Section (b): Internal conduit seal criteria is evaluated in PTN-Engineering    FPER-12-002.
Evaluation References                       Document ID PTN-FPER-12-002 Rev. 0 - Conduit Seal Compliance with NFPA 805 3.11.5 Electrical Raceway Fire   3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS).                         Compliance via The NRC provided an exemption for Thermo-Lag rated one hour Barrier Systems (ERFBS).         ERFBS required by Chapter 4 shall be capable of resisting the fire effects       Previous       and less (25 minutes). The revised Safety Evaluation states the of the hazards in the area. ERFBS shall be tested in accordance with and         Approval       following: The underlying purpose of Section III.G.2.a, Appendix shall meet the acceptance criteria of NRC Generic Letter 86-10,                                 R to 10 CFR Part 50, is to provide reasonable assurance that one Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier                         safe-shutdown train and associated circuits used to achieve and Systems Used to Separate Safe Shutdown Trains Within the Same Fire                              maintain safe-shutdown are free of fire damage. On the basis of Area." The ERFBS needs to adequately address the design requirements                            the staffs supporting safety evaluation of the licensee's and limitations of supports and intervening items and their impact on the                       submittals, the staff concludes that the exemption from the fire barrier system rating. The fire barrier system's ability to maintain the                   requirements of Section IIl.G.2.a of Appendix R to 10 CFR Part required nuclear safety circuits free of fire damage for a specific thermal                     50, for fire zones 79 (partial), 80 (partial), 82, 84 (partial), 85 exposure, barrier design, raceway size and type, cable size, fill, and type                     (partial), 88 (partial), 89 (partial), 91, 92, 105, and 117 as shall be demonstrated.                                                                          requested by the licensee, provides an adequate level of fire safety and presents no undue risk to public health and safety. In Exception No. 1: When the temperatures inside the fire barrier system                           addition, the staff concludes that the underlying purpose of the exceed the maximum temperature allowed by the acceptance criteria of                            rule is achieved. There have been no modifications to the fire Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire                         barrier that would invalidate this conclusion.
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 Turkey Point                                                      Run: 06/26/2012 11:05       Page: 46 of 48 Fire Safety Analysis Data Manager (4.1) 0 0                                                                                            0

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference         Requirements I Guidance                                                   Statement            Compliance Basis as specified by the AHJ at the time of acceptance.
References                       Document ID NRC-FPL Letter 5/5/99 - Issuance of a Revised Exemption and its Supporting Safety Evaluation for Fire Barriers in the Turbine Building Complies via         Evaluations of installed configurations are shown in the Engineering          Referenced Documents.
Evaluation References                       Document ID PTN-FPER-01-005 Rev. 0 - Evaluation of Fire Zone 117 Equivalent Barrier Protection PTN-FPER-06-002 Rev. 0 - Equivalency of Flame Retardant Coatings PTN-FPER-08-002 Rev. 0 - Acceptance Criteria for Thermo-Lag Leaching PTN-FPER-08-003 Rev. 0 - FHA for Conduit in Radioactive Pipeway PTN-FPER-89-021 Rev. 0- Thermo-Lag Fire Barrier Assembly Deviations PTN-FPER-89-025 Rev. 0 - Use of 3M FireDam Caulk with TSI Thermo-Lag Fire Barrier System PTN-FPER-91-003 Rev. 0 - Pull Box B-PB-5475 Thermo-Lag Fire Barrier PTN-FPER-91-004 Rev. 0 - Pull Box B-PB.5482 Thermo-Lag Fire Barrier PTN-FPER-93-001 Rev. 0 - Use of Topcoatings on Thermo-Lag Enclosures PTN-FPER-96-025 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the Unit 3 West Penetration Room PTN-FPER-96-026 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the ElevatorNestibule PTN-FPER-96-027 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the Electrical Equipment Room PTN-FPER-96-028 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 4160 Volt Switchgear Room 3B PTN-FPER-96-029 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 4B South DC Equipment Room PTN-FPER-96-030 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 3A North DC Equipment Room PTN-FPER-96-031 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 4A South Inverter DC Equipment Room PTN-FPER-96-032 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 3B North Inverter DC Equipment Room PTN-FPER-97-009 Rev. 0 - Impact of Cure Time on Thermo-Lag Fire Barrier Performance PTN-FPER-97-013 Rev. 1 - Evaluation of Turbine Lube Oil Fire PTN-FPER-97-015 Rev. 0 - Technical Evaluation for Turbine Building Fire Protection PTN-FPER-97-016 Rev. 0 - Evaluation of 3M Interam Fire Dam 150 Caulk as a Thermo-Lag Base Layer Joint Sealant PTN-FPER-97-023 Rev. 0 - Thermo-Lag Fire Barrier SystemUpgrades in the 4160 Volt Switchgear Room 3A PTN-FPER-97-024 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 480 Volt Load Center 3C & 3D Room PTN-FPER-97-025 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the 480 Volt Load Center 4C & 4D Room PTN-FPER-98-001 Rev. 0 - Evaluation of Thermo-Lag Configurations Not Bounded by Specificatin 3.21 PTN-FPER-98-006 Rev. 0 - Evaluation of Various Thermo-Lag Configurations As 25 Minute Fire Barriers PTN-FPER-98-011 Rev. 0 - Thermo-Lag Fire Barrier System Upgrades in the Unit 3 RHR Pump B Room Fire Safety Analysis Data Manager (4.1)                                                     Turkey Point                                                         Run: 06/26/2012 11:05   Page: 47 of 48

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference      Requirements I Guidance                                                Statement        Compliance Basis PTN-FPER-98-012 Rev. 0 - Evaluation of Thermo-Lag 770-1 For Use In 25 Minute Fire Zones Fire S 6   nalysis Data Manager (4.1)                                                  Turkey Point                                    Run: 06/26/2012 11:05 Page: 48 of 48 0
0

0 Florida Power & Light 0
Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 4
B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review 101 Pages Attached Page B-I Revision 00                                                                                                       Page B-1

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire 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 Methodology                         This section discusses a generic deterministic methodology and criteria that licensees can use to 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, Il.G.2 and llI.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 Applicable Alignment Statement                                 Alignment Basis                                                                    Reference Aligns                                              This is introductory information. The Turkey Point analysis utilized approaches    5610-M-722A Rev. 0 - Nuclear Safety Capability Fire Safe outlined in NEI 00-01 to identify the systems, components and cables needed        Shutdown Analysis Basis Document for evaluation of the plant's ability to achieve a safe and stable condition (in   561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe accordance with section 4.2.1.2 of this LAR Submittal) following a postulated      Shutdown Analysis (FSSA) fire. The analysis followed the general guidelines of this section.                NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                       Run: 06/26/2012 11:08     Page: 1 of 101 0                                                                                                0                                                                                          0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1 [A, Intro] Safe Shutdown Systems and Path  This section discusses the identification of systems available and necessary to perform the required safe shutdown functions. It also provides Development                                    information on the process for combining these systems into safe shutdown paths. Appendix R Section III.G.1 .a requires that the capability to achieve and maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section IIl.G.1 .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 Applicable Alignment Statement                           Alignment Basis                                                                    Reference Aligns                                         The Turkey Point analysis utilized approaches outlined in NEI 04-02 to identify   561 0-M-722A Rev. 0 [Attachments A, B, C, and D] - Nuclear Safety the systems, components and cables needed for evaluation of the plant's           Capability Fire Safe Shutdown Analysis Basis Document ability to achieve a safe and stable condition (in accordance with section        561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe 4.2.1.2 of this LAR Submittal) following a postulated fire. The analysis followed Shutdown Analysis (FSSA) the general guidelines of this section. The systems required to achieve hot and   561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown cold shutdown have been identified.                                               Analysis Essential Equipment List (EEL)
NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                      Run: 06/26/2012 11:08          Page: 2 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1 [B, Goals] Safe Shutdown Systems and      The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and components remains free of fire damage for a Path Development                              single fire in any single plant fire area. This goal is accomplished by determining those functions important to achieve and maintain hot shutdown.
Safe 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 Applicable Alignment Statement                           Alignment Basis                                                                       Reference Aligns                                         5610-M-722B and 5610-M-722A define the safe shutdown goals and                       5610-M-722A Rev. 0 [Attachments A and B] - Nuclear Safety functions. 5610-M-722A also identifies the safe shutdown success paths. All           Capability Fire Safe Shutdown Analysis Basis Document of the required safety functions are included in the analysis. 5610-M-723A list       561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe the required equipment.                                                               Shutdown Analysis (FSSA) 561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Analysis Essential Equipment List (EEL)
NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire       Analysis Data Manager (4.1)                                                          Turkey Point                                                        Run: 06/26/2012 11:08      Page: 3 of 101 0                                                                                            0

0                                                                                                                                                                                         0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1 [C, Spurious Operations] Safe Shutdown     In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated circuits with the potential for spurious equipment Systems and Path Development                  operation and/or loss of power source, and the common enclosure failures. Spurious operations/actuations can affect the accomplishment of the post
                                              -fire safe shutdown functions listed above. Typical examples of the effects of the spurious 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.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                       Reference Aligns                                        Cables identified as spurious concerns are identified in the safe shutdown           561 0-M-722A Rev. 0 [Attachments A, B, E, and F] - Nuclear Safety analysis, and are dispositioned in the same manner as other required safe            Capability Fire Safe Shutdown Analysis Basis Document shutdown cables. Components that are spurious actuation concerns are                  5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe identified as such on the Essential Equipment List. The multiple spurious            Shutdown Analysis (FSSA) operations (MSOs) PWROG generic list, determinations made by the plant's              5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown expert panel review, details about the location of an MSO logic representation       Analysis Essential Equipment List (EEL) and identification of analysis areas where specific MSOs occur are                    NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment documented in the Analysis Basis Document. Common enclosure concerns                  Database are also specifically addressed in the Analysis Basis Document.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                          Run: 06/26/2012 11:08       Page: 4 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1 Criteria / Assumptions                   The following criteria and assumptions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                 Reference NOT REQUIRED                                   This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                     Run: 06/26/2012 11:08     Page: 5 of 101 0                                                                                       0                                                                                        0

0                                                                                         0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.1 [GE BWR Paths]                         [BWR] GE Report GE-NE-T43-00002-00-01-RO1 entitled "Original Safe Shutdown Paths For The 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 II.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 Not Applicable Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                    Run: 06/26/2012 11:08     Page: 6 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.2 [SRVs / LP Systems]                    [BWR] GE Report GE-NE-T43-00002-00-03-RO1 provides a discussion on the BWR Owners' Group (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 II.G.1 and lII.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000.
Applicability Not Applicable Fire Safety Analysis Data Manager (4.1)                                                                                                                     Run: 06/26/2012 11:08   Page: 7 of 101 40 Turkey Point 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.3 [Pressurizer Heaters]                  [PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be 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.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                  Reference Aligns                                         Plant procedures, in conjunction with 0-ONOP-016.10 and 0-ONOP-105,             0-ONOP-016.10 Rev. 8 - Pre - Fire Plan Guidelines and Safe ensure the cooldown rate is controlled, adequate RCS inventory is maintained,    Shutdown Manual Actions and that the applicable pressure and temperature limits are not exceeded.        0-ONOP-105 Rev. 1 - Control Room Evacuation The procedures identify the availability or absence of pressurizer heaters and the method of providing feedwater and steam release.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                   Run: 06/26/2012 11:08       Page: 8 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.4 [Alternative Shutdown Capability]      The classification of shutdown capability as alternative shutdown is made independent of the 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 1ll.G.1 and Il.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 Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns                                         The plant's alternate safe shutdown capability was reviewed and approved by         0-ONOP-105 Rev. 1 - Control Room Evacuation the NRC in the April 16, 1984 SER. Procedure 0-ONOP-105 governs safe                FPL-NRC Letter L-83-516 - Alternate Safe Shutdown Capability shutdown of the plant for fires that require control room evacuation (including     Report 10 CFR 50 Appendix R Section III.L the Control Room, Cable Spreading Room, and North-South Breezeway).                 NRC-FPL Letter 4/16/84 - Alternate Safe Shutdown Capability - Fire Protection, Sections Ill.G.3 and III.L of Appendix R to 10 CFR 50, Turkey Point Plant Units 3 and 4 Comments Modifications required in the Alternate Shutdown Capability SER were implemented by PCM 83-154 for Unit 3 and PCM 83-155 for Unit 4 is Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point 0
Run: 06/26/2012 11:08 0
Page: 9 of 101

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.5 [Initial Conditions]                   At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) 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 Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns                                         These assumptions are clearly stated in the Analysis Basis Document.                5610-M-722A Rev. 0 [Attachments B section 3.2 and D section 4.5]
Specifically, the normal pre-fire condition for Component Selection will            - Nuclear Safety Capability Fire Safe Shutdown Analysis Basis consider the plant operating at 100% reactor power. Compliance Assessment           Document instructions further state that essential equipment not affected as a direct result of the fire is available.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                      Run: 06/26/2012 11:08    Page: 10 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                  NEI 00-01 Guidance 3.1.1.6 [Other Events in Conjunction with Fire] No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients need be considered in conjunction with the fire.
Applicability Applicable Alignment Statement                             Alignment Basis                                                                   Reference Aligns                                          This assumption is specifically stated in the plant's Safe Shutdown Analysis      5610-M-722A Rev. 0 [Attachment D section 4.4] - Nuclear Safety and the Compliance Assessment attachment of the Analysis Basis Document.          Capability Fire Safe Shutdown Analysis Basis Document The Compliance Assessment instruction states that design basis fires are assumed not to occur concurrently with non-fire related failures in safety systems, other accidents, or the most severe natural phenomena.
Run: 06/26/2012 11:08         Page: 11 of 101 0
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point 0                                                                                           0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.1.7 [ Offsite Power]                      For the case of redundant shutdown, offsite power may be credited ifdemonstrated to be free of fire 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.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                    Reference Aligns                                        The Analysis Basis Document's attachment on Compliance Assessment                  5610-M-722A Rev. 0 [Attachment D section 4.2] - Nuclear Safety states that: The availability or loss of off-site power will be assumed based      Capability Fire Safe Shutdown Analysis Basis Document upon the most conservative impact on the components under consideration.
Thus, the loss of off-site power cannot be credited such that the affected component can be assumed to fail to its desired safe shutdown position.
Credit will not be taken for off-site power in a specific fire area unless an analysis is performed to demonstrate the availability of off-site power following the worst case fire in the area in which all cables in the area are assumed damaged.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                    Run: 06/26/2012 11:08        Page: 12 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.8 [Safety-Related Equipment]             Post-fire safe shutdown systems and components are not required to be safety-related.
Applicability Applicable Alignment Statement                            Alignment Basis                                                                  Reference Aligns                                         This assumption is stated in section 4.9 of attachment D to the Analysis Basis   5610-M-722A Rev. 0 [Attachment B and attachment D section 4.9] -
Document as safe shutdown systems need not be designed to seismic                Nuclear Safety Capability Fire Safe Shutdown Analysis Basis category I criteria, single failure criteria, or cope with other plant accidents Document except those portions of these systems which interface with or impact existing safety systems.
Fire Safety Analysis Data Manager (4.1) is Turkey Point 0
Run: 06/26/2012 11:08 is Page: 13 of 101

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.9 [72 Hour Coping]                      The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor 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 Applicable Alignment Statement                           Alignment Basis                                                                    Reference Aligns                                        The plant (as indicated in Section 4.2.1.2 of this LAR Submittal) states that it   561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe will maintain the plant in a safe and stable condition indefinitely for all fires. Shutdown Analysis (FSSA)
Although the postulated fire scenario may not require it, the plant's analysis demonstrates the plant can be brought to cold shutdown within 72 hours for all fire areas if necessary to achieve this requirement.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                    Run: 06/26/2012 11:08       Page: 14 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.1.10 [Manual/Automatic Initiation of      Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown is acceptable Systems]                                      where permitted by current regulations or approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                   Reference Aligns                                        The section 5.6 of the Circuit Analysis attachment to the Analysis Basis          5610-M-722A Rev. 0 [Attachment C section 5.6] - Nuclear Safety Document states that:                                                             Capability Fire Safe Shutdown Analysis Basis Document Manual initiation of required systems from the control room or primary control station may be credited for this analysis. Automatic operation of specific components or functions within required systems is credited where appropriate and must be clearly identified for that component identifier. For such cases, all circuits for the automatic function must have been analyzed to remain free of fire damage. Some automatic system initiation will not be credited to align the component to its required position unless all circuits for this function have been analyzed. However, fire induced automatic initiation signals must be evaluated for the possibility of spurious component operation and their subsequent adverse inpact on safe shutdown.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                    Run: 06/26/2012 11:08      Page: 15 of 101 0                                                                                        0 0

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.1.11 [Multiple Affected Units]            Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and maintain safe shutdown for each affected unit must be demonstrated.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                       Reference Aligns                                        The ability of each unit to safely shutdown for a fire in any area within the plant   5610-M-722A Rev. 0 [Attachment D section 5.1] - Nuclear Safety is demonstrated in the safe shutdown analysis.                                       Capability Fire Safe Shutdown Analysis Basis Document 5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                        Run: 06/26/2012 11:08     Page: 16 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2 Shutdown Functions                      The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown. For additional information on BWR system selection, refer to GE Report GE-NE-T43-00002-00-01-RO1 entitled "Original Safe Shutdown Paths for the BWR."
Applicability Applicable Alignment Statement                           Alignment Basis                                                                Reference NOT REQUIRED                                  This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent subsections.
The GE information does not apply to either unit at Turkey Point.
Fire Safety Analysis Data Manager (4.1) 0                                                                                    Turkey Point is Run: 06/26/2012 11:08      Page: 17 of 101 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.2.1 Reactivity Control                    [BWR] Control Rod Drive System 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 There must be a method for ensuring that adequate shutdown margin is maintained by ensuring borated water is utilized for RCS makeup/charging.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                  Reference Aligns                                        The two credited sources of makeup water for post fire safe shutdown are the     5610-M-722A Rev. 0 [Sections 3.9.3 and 3.9.4] - Nuclear Safety Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST). The           Capability Fire Safe Shutdown Analysis Basis Document boric acid concentrations in each tank ensure that adequate shutdown margin      5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe will be maintained throughout the cooldown process. The borated water is         Shutdown Analysis (FSSA) added using the charging pumps via the normal charging path and/or the RCP       561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown seal injection lines.                                                            Analysis Essential Equipment List (EEL)
Fire Safety Analysis Data Manager (4.1)                                                     Turkey Point                                                       Run: 06/26/2012 11:08       Page: 18 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.2.2 Pressure Control Systems              The systems discussed in this section are examples of systems that can be used for pressure 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 Applicable Alignment Statement                           Alignment Basis                                                                  Reference Aligns                                        Makeup water is added using the charging pumps via the normal charging           5610-M-722A Rev. 0 [Sections 3.9 and 3.11] - Nuclear Safety path and/or the RCP seal injection lines. Auxiliary spray and pressurizer        Capability Fire Safe Shutdown Analysis Basis Document heaters are evaluated for availability to provide pressure control as well.      5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Should pressure reduction in excess of the capability of cool down or auxiliary  Shutdown Analysis (FSSA) spray be required, analysis demonstrates the availability of at least one PORV   561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown (using either instrument air or backup nitrogen bottles) for rapid               Analysis Essential Equipment List (EEL) depressurization. The control of spurious PORV, pressurizer heater, charging, and auxiliary spray initiation has also been analyzed as a part of pressure control.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                     Run: 06/26/2012 11:08 0                                                                                         0 Page: 19 of 101 0

0                                                                                         0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.3 Inventory Control                      [BWR] Systems selected for the inventory control function should be capable of supplying sufficient 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 Applicable Alignment Statement                           Alignment Basis                                                                   Reference Aligns                                        Manual initiation of the Chemical and Volume Control and Safety Injection         5610-M-722A Rev. 0 [Sections 3.9, 3.11, and 3.17] - Nuclear Safety Systems from the control room are available to provide RCS inventory control. Capability Fire Safe Shutdown Analysis Basis Document Spurious operation of components which may cause a loss of reactor coolant        561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe (e.g., letdown) or of RWST inventory is also evaluated.                           Shutdown Analysis (FSSA) 561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Analysis Essential Equipment List (EEL)
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                    Run: 06/26/2012 11:08      Page: 20 of 10 1

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.4 Decay Heat Removal                    [BWR] Systems selected for the decay heat removal function(s) should be capable of:
                                              - 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.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                   Reference Aligns                                        Although the plant's definition of safe and stable (see section 4.2.1.2 of this   561 0-M-722A Rev. 0 - Nuclear Safety Capability Fire Safe LAR Submittal) does not specifically require operators to achieve cold            Shutdown Analysis Basis Document shutdown, the ability to reach this state is demonstrated in the safe shutdown    561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe analysis for both units for a postulated fire in every area. Systems analyzed     Shutdown Analysis (FSSA) include Reactor Coolant, Main Steam, Auxiliary Feedwater, Standby Steam           5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Generator Feedwater, Residual Heat Removal, Component Cooling, and                Analysis Essential Equipment List (EEL)
Intake Cooling Water.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                       Run: 06/26/2012 11:08        Page: 21 of 101 0                                                                                          0                                                                                              0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.5 Process Monitoring                    The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1, 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 III.L.I. 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 Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns                                        The essential equipment list indicates that the minimum instrumentation listed      561 0-M-722A Rev. 0 - Nuclear Safety Capability Fire Safe in IN 84-09 is satisfied. Instrumentation includes pressure, temperature and        Shutdown Analysis Basis Document level for the RCS, pressure and level for the steam generators, flow for the        5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown auxiliary feedwater system, temperature for the component cooling water             Analysis Essential Equipment List (EEL) system, source range neutron flux, and level indication for all necessary sources of make-up water and diesel fuel oil.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                        Run: 06/26/2012 11:08      Page: 22 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.6 Support Systems                        [Blank Heading - No specific guidance]
Applicability Not Applicable Fire Safety Analysis Data Manager (4.1) 0                                                                                 Turkey Point 0
Run: 06/26/2012 11:08 Page: 23 of 101 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.2.6.1 Electrical Systems                  AC Distribution System 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 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. If the 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 Applicable Alignment Statement                           Alignment Basis                                                                        Reference Aligns                                        Analysis evaluates availability of power from off-site and on-site sources in          5610-M-722A Rev. 0 [Sections 3.12 and 3.13] - Nuclear Safety III.G.1 and III.G.2 areas (though credit is not taken for a fire causing a loss of     Capability Fire Safe Shutdown Analysis Basis Document off-site power if such loss would be beneficial). III.G.3 areas only analyze the       5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe availability of on-site power to support safe shutdown operations. Power is            Shutdown Analysis (FSSA) evaluated for loads at the 4kV, 480V, and 120V AC levels and at the 125V DC            5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown level. DC distribution requires the availability of the associated station battery    Analysis Essential Equipment List (EEL) and a battery charger for availability beyond two hours.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                        Run: 06/26/2012 11:08       Page: 24 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.6.2 Cooling Systems [Main Section]      Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations. Typical uses include:
                                               - RHR/SDC/DH Heat Exchanger cooling water
                                               - Safe shutdown pump cooling (seal coolers, oil coolers)
                                               - Diesel generator cooling
                                               - HVAC system cooling water Applicability Applicable Alignment Statement                           Alignment Basis                                                                Reference Aligns                                        The Component Cooling Water (CCW) and Intake Cooling Water (ICW)               5610-M-722A Rev. 0 [Sections 3.8, 3.13, 3.14, and 3.15] - Nuclear systems have been analyzed. Cooling systems for the Emergency Diesel           Safety Capability Fire Safe Shutdown Analysis Basis Document Generators, Diesel Air Compressors, HVAC systems outside of containment,      5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe and Diesel Fire Pump are all self-contained and do not require external water Shutdown Analysis (FSSA) cooling.                                                                      5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Analysis Essential Equipment List (EEL) is Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point 0
Run: 06/26/2012 11:08     Page: 25 of 101
                                                                                                                                                                                     .0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.2.6.2 [A] Cooling Systems [HVAC]           HVAC Systems 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 Applicable Alignment Statement                           Alignment Basis                                                                   Reference Aligns                                        Ventilation for the EDG Building, Auxiliary Building, Switchgear and Load         561 0-M-722A Rev. 0 [Sections 3.1, 3.2, and 3.13] - Nuclear Safety Center Rooms, and Control Building (including Control Room, Cable                Capability Fire Safe Shutdown Analysis Basis Document Spreading Room, and DC Equipment Rooms) is analyzed. Containment                 561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Cooling is also evaluated.                                                        Shutdown Analysis (FSSA) 5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Analysis Essential Equipment List (EEL)
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                      Run: 06/26/2012 11:08          Page: 26 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.1.3 Methodology for Shutdown System         Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown 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 Applicable Alignment Statement                           Alignment Basis                                                                        Reference NOT REQUIRED                                  This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent subsections.
      .0 Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                     Run: 06/26/2012 11:08    Page: 27 of 101 0                                                                                        0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.3.1 Identify safe shutdown functions       Review available documentation to obtain an understanding of the available plant systems and the 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 Applicable Alignment Statement                            Alignment Basis                                                                  Reference Aligns                                         The PTN safe shutdown systems were chosen to provide redundant means of          561 0-M-722A Rev. 0 [Attachment B, section 5.1]- Nuclear Safety accomplishing the required safe shutdown functions. Development of the PTN       Capability Fire Safe Shutdown Analysis Basis Document safe shutdown equipment list and logics was accomplished by reviewing operating instructions and procedures (emergency, off-normal and normal),
fire hazards analysis, P&IDs, system descriptions, design basis documents, single-line diagrams, elementary diagrams, instrument loop diagrams, and logic diagrams.
Fire Safety Analysis Data Manager (4. 1)                                                      Turkey Point                                                    Run: 06/26/2012 11:08     Page: 28 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.3.2 Identify Combinations of Systems that  Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of systems capable of achieving the safe shutdown Satisfy Each Safe Shutdown Function            functions of reactivity control, pressure control, inventory control, decay heat removal, process monitoring, and support systems such as electrical and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other 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.
Applicability Applicable Alignment Statement                            Alignment Basis                                                                      Reference Aligns                                         The available combinations of systems have been identified and are shown on          5610-M-722A Rev. 0 [Attachment A and attachment B section 5.2] -
the safe shutdown logic diagrams developed as part of the validation                 Nuclear Safety Capability Fire Safe Shutdown Analysis Basis project. These logics include all spurious operation concerns identified at the      Document plant. Power supply issues are identified through dependency relationships in        NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment the compliance assessment database.                                                  Database Fire Safety Analysis Data Manager (4.1) 0 Turkey Point is Run: 06/26/2012 11:08 is Page: 29 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.3.3 Define Combinations of Systems for     Select combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set of systems as a Each Safe Shutdown Path                        safe shutdown path. In many cases, safe shutdown paths may be defined on a divisional basis since the availability of electrical power and other support systems must be demonstrated for each path.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns with Intent                             The safe shutdown logics show the combinations of systems available, but            561 O-M-722A Rev. 0 [Attachment A and attachment B section 5.2] -
specific "paths" are not always identified. Where applicable, specific train,       Nuclear Safety Capability Fire Safe Shutdown Analysis Basis electrical division, and other support systems are evident in the safe shutdown     Document logic diagrams.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                        Run: 06/26/2012 11:08     Page: 30 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.1.3.4 Assign Shutdown Paths to Each         Assign a path designation to each combination of systems. The path will serve to document the combination of systems relied upon for safe Combination of Systems                         shutdown in each fire area. Refer to Attachment 1 to this document (NEI 00-01) for an example of a table illustrating how to document the various combinations of systems for selected shutdown paths.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                  Reference Aligns with Intent                             Path designations are not assigned to each possible combination of systems      5610-M-722A Rev. 0 [Attachments A, D (Section 5.1)] - Nuclear available to achieve safe shutdown. Rather, a specific combination of systems   Safety Capability Fire Safe Shutdown Analysis Basis Document capable of achieving each performance goal has been identified for each unit    NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment for each fire area. Within these credited system are components and systems      Database relied upon to perform the shutdown function and whose spurious operation could adversely affect that system's ability to achieve its safe shutdown function.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                    Run: 06/26/2012 11:08     Page: 31 of 101 is                                                                                        0 40

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2 Safe Shutdown Equipment Selection         The previous section described the methodology for selecting the systems and paths necessary to 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 Applicable Alignment Statement                           Alignment Basis                                                                   Reference NOT REQUIRED                                  This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent sub-sections.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                     Run: 06/26/2012 11:08      Page: 32 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1 Criteria / Assumptions                  Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:
Applicability Applicable Alignment Statement                           Alignment Basis                                                                 Reference NOT REQUIRED                                  This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent subsections.
Fire Safety Analysis Data Manager (4.1) is Turkey Point 0
Run: 06/26/2012 11:08 is Page: 33 of 101

0                                                                                         0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.2.1.1 [Primary/ Secondary Components]       3.2.1.1 Safe shutdown equipment can be divided into two categories. Equipment may be categorized as (1) primary components or (2) secondary components. Typically, the following types of 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 interlock 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 Applicable Alignment Statement                           Alignment Basis                                                                       Reference Aligns                                        In general, primary components are included in the Essential Equipment List           5610-M-722A Rev. 0 [Attachments B (5.1.3, 5.1.4), C (6.1.3, 6.1.6)]
(EEL) and are flagged as safe shutdown components. Secondary                          - Nuclear Safety Capability Fire Safe Shutdown Analysis Basis components are generally not identified in the EEL, unless doing so would             Document enhance the clarity of the analysis. From a circuit analysis standpoint, any          NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment cables required for secondary components would be captured in the analysis            Database for the applicable primary component.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                         Run: 06/26/2012 11:08       Page: 34 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.2.1.2 [Fire Damage to Mechanical            3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundary or Components (not electrically supervised)]     safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in this assumption). Fire damage should be evaluated with respect to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.
Applicability Not Applicable Alignment Statement                           Alignment Basis                                                                     Reference Aligns with Intent                            It is assumed in the analysis that mechanical components like those identified      5610-M-722A Rev. 0 [Attachment D section 4.6] - Nuclear Safety will not suffer fire damage to the extent that they will not be able to perform the Capability Fire Safe Shutdown Analysis Basis Document credited safe shutdown function. The ability to manually operate valves within      PTN-FPER-11-002 Rev. 0 - NFPA 805 Recovery Action Feasibility the fire area (when necessary) is evaluated as part of the manual action            Evaluation feasibility review. Brazed components were assumed not to fail as a result of a fire in this analysis.
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point is Run: 06/26/2012 11:08 is Page: 35 of 101

                                                                                                 *0                                                                                   0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1.3 [Manual Valve Positions]              Assume that manual valves are in their normal position as shown on P&lDs or in the plant operating procedures.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                Reference Aligns                                        All credited components are assumed to be in their normal positions (as        5610-M-722A Rev. 0 [Attachment B section 3.1]- Nuclear Safety identified on the P&lDs or in system operating procedures) at the time of the Capability Fire Safe Shutdown Analysis Basis Document fire.
Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                   Run: 06/26/2012 11:08  Page: 36 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1.4 [Check Valves]                        Assume that a check valve closes in the direction of potential flow diversion and seats properly with 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.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                     Reference Aligns                                         Check valves are assumed to be leak tight when credited as system                  5610-M-722A Rev. 0 [Attachment B section 3.3] - Nuclear Safety boundaries.                                                                        Capability Fire Safe Shutdown Analysis Basis Document 0
Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                         Run: 06/26/2012 11:08       Page: 37 of 101 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1.5 [Instrument Failures]                  Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow 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 Applicable Alignment Statement                           Alignment Basis                                                                   Reference Aligns                                        Damage to instrument cables is assumed to fail the instrument in the least        561 0-M-722A Rev. 0 [Attachment C section 6.3] - Nuclear Safety desirable state.                                                                  Capability Fire Safe Shutdown Analysis Basis Document Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                      Run: 06/26/2012 11:08       Page: 38 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1.6 [Spurious Components]                 Identify equipment that could spuriously operate or mal-operate and impact the performance of equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                  Reference Aligns                                        Spurious operation was considered in identification of SSEL components. No       561 0-M-722A Rev. 0 [Attachment B section 4.2 and att. C section initial effort was made to 'bin' the types of potential spuriously operating     5.8] - Nuclear Safety Capability Fire Safe Shutdown Analysis Basis components or their cables (per RIS 2004-03). Spurious operation was             Document considered later during compliance assessment when circuit analysis was performed to determine if potential spurious operation was a concern requiring mitigating actions or other compliance strategies.
Fire Safety Analysis Data Manager (4.1) 0                                                                                          is Turkey Point                                                   Run: 06/26/2012 11:08    Page: 39 of 101 0

0                                                                                          0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.1.7 [Instrument Tubing]                    Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a 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 Applicable Alignment Statement                            Alignment Basis                                                                     Reference Aligns                                         The potential effects of fire damage to instrument tubing has been considered.      561 0-M-722A Rev. 0 [Section 3.15] - Nuclear Safety Capability Fire The only safe shutdown component with instrument tubing which crosses               Safe Shutdown Analysis Basis Document analysis area boundaries are the atmospheric dump valves (ADVs - going              PTN-FPER-10-009 Rev. 1 [RFI-0148] - SSA Upgrade Design from the steam platforms to the control room for control functions). The tubing     Support Input was evaluated for loss of control function, and to confirm that localized effects will not adversely affect control signals or indications that would affect post-fire shutdown capability.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                     Run: 06/26/2012 11:08     Page: 40 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                NEI 00-01 Guidance 3.2.2 Methodology for Equipment Selection     Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown 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]
Applicability Applicable Alignment Statement                           Alignment Basis                                                                        Reference NOT REQUIRED                                  This paragraph provides introductory information but contains no specific guidance for comparison. Discussion is provided in subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                                                                                         Run: 06/26/2012 11:08 Page: 41 of 101 0                                                                                         0 Turkey Point 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.2.1 Identify the System Flow Path for Each Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown path. Refer to Attachment 2 for an Shutdown Path                                  example of an annotated P&ID illustrating this concept.
Applicability Applicable Alignment Statement                            Alignment Basis                                                                  Reference Aligns                                         P&IDs were marked up for one-time use only when validating the safe              5610-M-722A Rev. 0 [Attachment B section 5.1.2] - Nuclear Safety shutdown equipment list and developing system logics.                            Capability Fire Safe Shutdown Analysis Basis Document Fire Safety Analysis Data Manager (4. 1)                                                      Turkey Point                                                    Run: 06/26/2012 11:08     Page: 42 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.2.2 Identify the Equipment in Each Safe    Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to assure that all equipment in each system's flow Shutdown System Flow Path Including            path has been identified. Assure that any equipment that could spuriously operate and adversely affect the desired system function(s) is also Equipment That May Spuriously Operate and      identified. If additional systems are identified which are necessary for the operation of the safe shutdown system under review, include these as Affect System Operation                       systems required for safe shutdown. Designate these new systems with the same safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1).
Applicability Applicable Alignment Statement                           Alignment Basis                                                                       Reference Aligns                                        All electrically supervised equipment in each credited flow path were identified      561 0-M-722A Rev. 0 [Attachment B section 5.1]- Nuclear Safety as safe shutdown components. All manual valves and dampers requiring                 Capability Fire Safe Shutdown Analysis Basis Document operation or which make up system boundaries (including check valves) were           NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment also identified as safe shutdown components. Power supplies for all active           Database components were also identified as safe shutdown components. Components whose spurious operation could adversely affect a system's ability to perform its safe shutdown function were identified as safe shutdown components and associated with that system in the NFPA 805 Compliance Assessment Database.
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point 0
Run: 06/26/2012 11:08 is Page: 43 of 101

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.2.3 Develop a List of Safe Shutdown        Prepare a table listing the equipment identified for each system and the shutdown path that it supports. Identify any valves or other equipment that Equipment and Assign the Corresponding         could spuriously operate and impact the operation of that safe shutdown system. Assign the safe shutdown path for the affected system to this System and Safe Shutdown Path(s)               equipment. During the cable selection phase, identify additional equipment required to support the safe shutdown function of the path (e.g.. electrical Designation to Each.                           distribution system equipment). Include this additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe shutdown and it documents various equipment-related attributes used in the analysis.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                    Reference Aligns                                         The table identifying the system and shutdown path each component supports        5610-M-722A Rev. 0 [Attachment B, attachment C section 6.0, & att.
was created as part of the Compliance Assessment database. All required           E] - Nuclear Safety Capability Fire Safe Shutdown Analysis Basis supporting equipment and spurious components have been identified.                Document 5610-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown Analysis Essential Equipment List (EEL)
NICAD Rev. 1.0a [Shutdown Systems Table] - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                      Run: 06/26/2012 11:08     Page: 44 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.2.4 Identify Equipment Information         Collect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for the equipment. In order to facilitate Required for the Safe Shutdown Analysis        the analysis, tabulate this data for each piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL.
Examples of related equipment data should include the equipment type, equipment 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 Applicable Alignment Statement                            Alignment Basis                                                                   Reference Aligns                                         The normal, required (HSD and CSD), and failed (air and electrical)               561 0-M-722A Rev. 0 [Attachment B section 4.1]- Nuclear Safety information is contained in the NFPA 805 Database in the Essential                Capability Fire Safe Shutdown Analysis Basis Document Equipment List (EEL). The list also identifies high/low pressure interface and    561 0-M-723A Rev. 0 - Nuclear Safety Capability Fire Shutdown spurious operation concerns.                                                      Analysis Essential Equipment List (EEL)
NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                       Run: 06/26/2012 11:08        Page: 45 of 101 0                                                                                         0                                                                                              0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.2.2.5 Identify Dependencies Between          In the process of defining equipment and cables for safe shutdown, identify additional supporting equipment such as electrical power and interlocked Equipment, Supporting Equipment, Safe         equipment. As an aid in assessing identified impacts to safe shutdown, consider modeling the dependency between equipment within each safe Shutdown Systems and Safe Shutdown Paths. shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these relationships.
Applicability Applicable Alignment Statement                           Alignment Basis                                                                     Reference Aligns                                        The required support equipment, including electrical power has been                561 O-M-722A Rev. 0 [Att. A, att. B section 5.2, & att. C section identified, and the relationships are captured either in the database or on the    6.1.6] - Nuclear Safety Capability Fire Safe Shutdown Analysis safe shutdown logic diagrams.                                                      Basis Document NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                      Run: 06/26/2012 11:08        Page: 46 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety 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 fire-induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified.
This situation could occur ifthe 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 Cable Selection and                  This section provides industry guidance on the recommended methodology and criteria for selecting safe shutdown cables and determining their Location                                               potential impact on equipment required for achieving 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 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 Applicable Alignment Statement                                    Alignment Basis                                                                     Reference NOT REQUIRED                                           This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1) 0                                                                                              Turkey Point                                                         Run: 06/26/2012 11:08       Page: 47 of 10 1 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1 Criteria / Assumptions                     To identify an impact to safe shutdown equipment based on cable routing, the equipment must have 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 Applicable Alignment Statement                              Alignment Basis                                                                  Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                  Run: 06/26/2012 11:08          Page: 48 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.1 [Cable Selection]                        The list of cables whose failure could impact the operation of a piece of safe shutdown equipment 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 Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           The cable selection and circuit analysis instruction discusses in detail the          5610-M-722A Rev. 0 [Attachment C sections 5 and 6] - Nuclear circuit analysis methodology in use at Turkey Point. The safe shutdown                Safety Capability Fire Safe Shutdown Analysis Basis Document database contains the information developed during the circuit analysis. This         5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe includes linking power supply dependencies to ensure components that                  Shutdown Analysis (FSSA) require power have that power supply free of fire damage.                             NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database 0
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                        Run: 06/26/2012 11:08        Page: 49 of 101 0                                                                                              0

0                                                                                        0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.2 [Cables Affecting Multiple Components]   In cases where the failure (including spurious actuations) of a single cable could impact more than one piece of safe shutdown equipment, include the cable with each piece of safe shutdown equipment.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                     Reference Aligns with Intent                               Although this specific guidance from this section of NEI 00-01 is not repeated      561 0-M-722A Rev. 0 [Attachment C section 6] - Nuclear Safety in the cable selection and circuit analysis instruction, it was followed in         Capability Fire Safe Shutdown Analysis Basis Document practice. The procedure requires that ALL cables that could adversely affect the ability of the component to perform its safe shutdown function be identified.
Nothing in the procedure or database prevents assigning a safe shutdown cable to more than one component. The procedure also provides guidance on whether an interlocked or off-scheme cable be identified with the component being analyzed or assigned to its own (or another) safe shutdown component and the interrelationship shown on the safe shutdown system logic diagrams.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                     Run: 06/26/2012 11:08     Page: 50 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.3 [Isolation Devices]                      Electrical devices such as relays, switches and signal resistor units are considered to be acceptable 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 Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           The circuit analysis and cable selection procedure requires identification            561 0-M-722A Rev. 0 [Attachment C sections 6.1.1O.b) and c)] -
of suitable isolation devices before a cable can be considered "Not Required"         Nuclear Safety Capability Fire Safe Shutdown Analysis Basis for safe shutdown. Optical isolators and other isolation circuits are evaluated       Document for their capability to isolate cables to non-credited functions such as the process computer or annunciators on all circuits, including instrument loops.
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                       Run: 06/26/2012 11:08 is Page: 51 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.4 [Identify "Not Required" Cables]         Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                    Reference Aligns                                           Cables which do not impact a component's ability to perform its safe shutdown      561 0-M-722A Rev. 0 [Attachment C section 6.1.9] - Nuclear Safety function and are properly isolated from the required portion of the component's    Capability Fire Safe Shutdown Analysis Basis Document circuit are identified as "Not Required" in the safe shutdown database and are     NICAD Rev. 1.0a [Circuit Analysis Worksheets] - PTN NFPA 805 not analyzed further for that component.                                           Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                      Run: 06/26/2012 11:08       Page: 52 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.5 [Identification of Power Supplies]       For each circuit requiring power to perform its safe shutdown function, identify the cable supplying 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                     Reference Aligns                                           For individual components, power supply cables were identified up to the            561 O-M-722A Rev. 0 [Attachment C sections 6.7 and 6.8] - Nuclear immediate upstream power source. The power source was then added to the             Safety Capability Fire Safe Shutdown Analysis Basis Document Essential Equipment List as a safe shutdown component, and circuit analysis was performed on the power supply. This was performed back to the EDGs and off-site power transformers and their switchyard breakers.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                      Run: 06/26/2012 11:08      Page: 53 of 101 0

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.6 [ESFAS Initiation]                       The automatic initiation logics for the credited post-fire safe shutdown systems are not required to support safe shutdown. Each system can be controlled manually by operator actuation in the main control room or emergency control station. If operator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, ifnot 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 Applicable Alignment Statement                              Alignment Basis                                                                         Reference Aligns                                           Spurious safeguards actuation was evaluated for its ability to cause spurious           561 0-M-722A Rev. 0 [Attachments A 3-SSS21&4 -SSS21 & C component realignments which could impact the safe shutdown success                     section 5.6] - Nuclear Safety Capability Fire Safe Shutdown Analysis path. Where a spurious ESFAS signal could occur, operator actions within or             Basis Document outside the control room were identified as necessary. All actions identified           PTN-FPER-11-002 Rev. 0 [Sections 1 and 6] - NFPA 805 Recovery were evaluated for feasibility and reliability as required by regulations.              Action Feasibility Evaluation Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                         Run: 06/26/2012 11:08      Page: 54 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.1.7 [Circuit Coordination]                   Cabling for the electrical distribution system is a concern for those breakers that feed associated 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 the load center would also be necessary to support the load center.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                      Reference Aligns with Intent                               The circuit analysis procedure assumes coordination exists. In the SER dated         5610-M-722A Rev. 0 [Attachment C section 3.2] - Nuclear Safety April 16, 1984, the staff stated (based on the FPL submittal) that because           Capability Fire Safe Shutdown Analysis Basis Document coordination of circuit protective devices was part of the original electrical       NRC-FPL Letter 4/16/84 [Section II.B.1, page 4] - Alternate Safe system design, associated circuits by common power supply should not exist.          Shutdown Capability - Fire Protection, Sections III.G.3 and III.L of However, this review was for the plant's alternate shutdown capability, and all      Appendix R to 10 CFR 50, Turkey Point Plant Units 3 and 4 power supplies being credited in the re-validation may not have been evaluated.
ITEM FOR IMPLEMENTATION (see PTN NFPA 805 LAR Attachment S Table S-2 Items 19 and 20):
* Perform modifications as necessary to achieve coordination on DP412A, 4DP86, and 4DP87 (only NSCA power supplies without demonstrated coordination)
This task will be performed by FPL.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                      Run: 06/26/2012 11:08        Page: 55 of 101 0                                                                                                                                                                                          0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.2 Associated Circuit Cables                  Appendix R, Section III.G.2, requires that separation features be provided for equipment and cables, including associated nonsafety 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 Applicable Alignment Statement                              Alignment Basis                                                                    Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                       Run: 06/2612012 11:08     Page: 56 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.2 [A] Associated Circuit Cables - Cables     Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose failure could cause the spurious actuation/mal-Whose Failure May Cause Spurious Actuations      operation of equipment whose operation could affect safe shutdown. These cables are identified in Section 3.3.3 together with the remaining safe shutdown cables required to support control and operation of the equipment.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                  Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                     Run: 06/26/2012 11:08      Page: 57 of 101 0                                                                                          0 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.2 [B] Associated Circuit Cables - Common     The concern for the common power source associated circuits is the loss of a safe shutdown power source due to inadequate breaker/fuse Power Source Cables                              coordination. In the case of a fire-induced cable failure on a non-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordination 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 Applicable Alignment Statement                              Alignment Basis                                                                    Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                      Run: 06/26/2012 11:08    Page: 58 of 10 1

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.2 [C] Associated Circuit Cables - Common     The concern with common enclosure associated circuits is fire damage to a cable whose failure could propagate to other safe shutdown cables in the Enclosure Cables                                 same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result 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 Applicable Alignment Statement                              Alignment Basis                                                                       Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point 0
Run: 06/26/2012 11:08 is Page: 59 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Reft                                   NEI 00-01 Guidance 3.3.3 Methodology for Cable Selection and        Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables necessary for performing a post-fire safe shutdown Location                                         analysis.
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 3-4]
Applicability Applicable Alignment Statement                              Alignment Basis                                                                        Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                       Run: 06/2612012 11:08       Page: 60 of 10 1

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                    NEI 00-01 Guidance 3.3.3.1 Identify Circuits Required for the       For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the following documentation to Operation of the Safe Shutdown Equipment         identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation of each piece of equipment:
                                                 - 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                      Reference Aligns with Intent                               The analyst obtained the applicable schematic diagrams and other available           5610-M-722A Rev. 0 [Attachment C sections 3.2 and 6.1] - Nuclear design input documentation to perform an adequate circuit analysis of the            Safety Capability Fire Safe Shutdown Analysis Basis Document component's ability to perform its safe shutdown function (including availability of power supplies) as required by the circuit analysis and cable selection instruction.
is Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point is Run: 06/26/2012 11:08      Page: 61 of 101 0

0                                                                                          0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.3.2 Identify Interlocked Circuits and Cables In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and equipment. Assign to the equipment Whose Spurious Operation or Mal-operation        any cables for interlocked circuits that can affect the equipment.
Could Affect Shutdown                            While investigating the interlocked circuits, additional equipment or power sources may be discovered. Include these interlocked equipment or power sources in the safe shutdown equipment list (refer to Figure 3-3) ifthey can impact the operation of the equipment under consideration.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           The circuit analysis and cable selection instruction provides detailed guidance       5610-M-722A Rev. 0 [Attachment C section 6.1.5] - Nuclear Safety concerning the treatment of interlocked circuits and when the analyst should          Capability Fire Safe Shutdown Analysis Basis Document consider creating a separate equipment identifier for them versus listing the interlock cables against the primary component.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                       Run: 06/26/2012 11:08     Page: 62 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.3.3.3 Assign Cables to the Safe Shutdown       Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result in maloperation of each piece of safe Equipment                                        shutdown 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                     Reference Aligns with Intent                               The required cables for each component have been identified and tabulated in        5610-M-722A Rev. 0 [Attachment C sections 3.2, 5 and 6] - Nuclear a relational database. The circuit analysis and cable selection instruction         Safety Capability Fire Safe Shutdown Analysis Basis Document assumes coordination exists. In the SER dated April 16, 1984, the staff stated      NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment (based on the FPL submittal) that because coordination of circuit protective        Database devices was part of the original electrical system design, associated circuits by   NRC-FPL Letter 4/16/84 [Section Il.B.1, pg. 4] -Alternate Safe common power supply should not exist. However, this review was for the              Shutdown Capability - Fire Protection, Sections III.G.3 and IIl.L of plant's alternate shutdown capability, and all power supplies being credited in     Appendix R to 10 CFR 50, Turkey Point Plant Units 3 and 4 the re-validation may not have been evaluated.
ITEM FOR IMPLEMENTATION (see PTN NFPA 805 LAR Attachment S Table S-2 Items 19 and 20):
* Perform modifications as necessary to achieve coordination on DP412A, 4DP86, and 4DP87 (only NSCA power supplies without demonstrated coordination)
This task will be performed by FPL.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                       Run: 06/26/2012 11:08      Page: 63 of 10 1 0                                                                                           is                                                                                            0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5 Circuit Analysis and Evaluation              This section on circuit analysis provides information on the potential impact of fire on circuits used to 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 Applicable Alignment Statement                              Alignment Basis                                                                        Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                         Run: 06/26/2012 11:08     Page: 64 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1 Criteria / Assumptions                     Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                     Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                        Run: 06/26/2012 11:08 Page: 65 of 101 0                                                                                                                                                                  0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.1 [Circuit Failure Types and Impact]       Consider the following circuit failure types on each conductor of each unprotected safe shutdown 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 Applicable Alignment Statement                             Alignment Basis                                                                        Reference Aligns                                           Circuit analysis was performed and has been documented in the NFPA 805                561 0-M-722A Rev. 0 [Attachment C Sect, 5 & 6 and attachment D Transition Database prior to performing safe shutdown compliance                      Sect. 5] - Nuclear Safety Capability Fire Safe Shutdown Analysis assessment validation. Thus, only those cables that have been previously              Basis Document determined to adversely affect the ability of the component to perform its safe shutdown function have been identified as required cables. This includes consideration of hot shorts, open circuits, and short circuits to ground.
Fire Safety Analysis Data Manager (4.1)                                                             Turkey Point                                                      Run: 06/26/2012 11:08      Page: 66 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.2 [Circuit Contacts and Operational        Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal mode/position of the safe shutdown equipment as shown Modes]                                           on the schematic drawings. The analyst 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 Applicable Alignment Statement                              Alignment Basis                                                                     Reference Aligns with Intent                               The analysis assumes components are in their normal operating position or           561 0-M-722A Rev. 0 [Att. B sect. 3.1, & att. C sect. 5.9, 5.10, &
condition, and this is understood to include contacts within the components'        5.11] - Nuclear Safety Capability Fire Safe Shutdown Analysis Basis electrical circuits. Specific instructions are provided concerning local control    Document station isolation switches and transfer switches. Analysis also evaluates potential for contacts to reposition during the course of a normal shutdown to CSD or other impacts from the selected safe shutdown success path.
0                                                                                              0 Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                     Run: 06/26/2012 11:08        Page: 67 of 101 40

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.3 [Duration of Circuit Failures]           Assume that circuit failure types resulting in spurious operations exist until action has been taken to 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           Mitigation or clearing of spurious signals was only assumed to occur when             5610-M-722A Rev. 0 [Attachments C and D] - Nuclear Safety positive operator action is taken, except where circuit timing devices would          Capability Fire Safe Shutdown Analysis Basis Document clear the signal itself. Hot shorts resulting in spurious actuations are not assumed to "self-mitigate" after a limited period of time.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                          Run: 06/26/2012 11:08      Page: 68 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.4 [Cable Failure Configurations]           When both trains are in the same fire area outside of primary containment, all cables that do not meet the separation requirements of Section II.G.2 are assumed to fail in their worst case configuration.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                    Reference Aligns                                           The analysis and supporting procedures for the revalidation assume that all        5610-M-722A Rev. 0 [Attachments C (Sect. 3.1)and D (Sect. 4.2 cables in the area being analyzed fail to the worst case position. This includes   and 5.0)] - Nuclear Safety Capability Fire Safe Shutdown Analysis taking into account that power or control air may or may not be available, such    Basis Document that the analysis does not credit a potential loss of power to dismiss a potential 5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe spurious actuation.                                                                Shutdown Analysis (FSSA)
Run: 06/26/2012 11:08     Page: 69 of 101 0
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point 0                                                                                           0

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.5 [A, Circuit Failure Risk Assessment      The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to identify any potential combinations of spurious Guidance]                                        operations with higher risk significance. Bin 1 failures should also be the focus of the analysis; however, NRC has indicated that other types of failures 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 Applicable Alignment Statement                              Alignment Basis                                                                     Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                       Run: 06/26/2012 11:08      Page: 70 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.1.5 [B, Cable Failure Modes]                 For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the same cable is the most common mode of failure.
This is often referred to as "intra-cable shorting." It 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 spurious 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 startlstop/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 Applicable Alignment Statement                              Alignment Basis                                                                        Reference Aligns with Intent                               Cable selection considered any and all hot shorts, regardless of whether these         5610-M-722A Rev. 0 [Attachment C] - Nuclear Safety Capability Fire hot shorts were inter-cable or intra-cable. No restriction was placed on the           Safe Shutdown Analysis Basis Document number of cables or hot shorts required to fail the component/circuit. No credit was taken for relative risk due to conductor combinations available for a fault.
All cables were conservatively assumed to be Thermo-plastic. Cables were not excluded because of being instrumentation cables or because they were part of a DC circuit requiring multiple hot shorts of proper polarity.
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point 0
Run: 06/26/2012 11:08 is Page: 71 of 101

0                                                                                                                                                                                            0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                    NEI 00-01 Guidance 3.5.2 Types of Circuit Failures                  Appendix R requires that nuclear power plants must be designed to prevent exposure fires from 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 Applicable Alignment Statement                              Alignment Basis                                                                      Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                          Run: 06/26/2012 11:08        Page: 72 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.1 Circuit Failures Due to an Open Circuit  This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of 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:
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 Applicable Alignment Statement                              Alignment Basis                                                                         Reference Aligns                                           The Turkey Point SSA does consider open circuits. This section provides                 5610-M-722A Rev. 0 [Attachment C section 5.0] - Nuclear Safety information related to the effects of an open circuit on different types of typical     Capability Fire Safe Shutdown Analysis Basis Document circuits including CT secondary circuits.                                               5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
PTN-ENG-SEES-1 1-111 - Evaluation for Open Circuited Current Transformers Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                            Run: 06/26/2012 11:08     Page: 73 of 101 0                                                                                            0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.2 Circuit Failures Due to a Short-to-      This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown equipment. A short-to-ground is a fire-Ground [A, General]                              induced breakdown of a cable insulation system resulting in the potential on the conductor being applied to ground potential. A short-to-ground.can 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 Applicable Alignment Statement                              Alignment Basis                                                                      Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                          Run: 06/26/2012 11:08      Page: 74 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.2 Circuit Failures Due to a Short-to-      This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown equipment. A short-to-ground is a fire-Ground [B, Grounded Circuits]                    induced breakdown of a cable insulation system resulting in the potential on the conductor being applied to ground potential. A short-to-ground can 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 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 Applicable Alignment Statement                              Alignment Basis                                                                        Reference Aligns                                           The Turkey Point SSA does consider shorts to ground. This section provides             561 0-M-722A Rev. 0 [Attachment C section 5.0] - Nuclear Safety information related to the effects of a short to ground on different types of          Capability Fire Safe Shutdown Analysis Basis Document typical grounded circuits.                                                             561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                         Run: 06/26/2012 11:08 is Page: 75 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.2 Circuit Failures Due to a Short-to-      Short-to-Ground on Ungrounded Circuits Ground [C, Ungrounded Circuits]
In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the 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.
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 circuitor 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.
S.hort-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 Applicable Alignment Statement                              Alignment Basis                                                                        Reference Aligns                                           The Turkey Point SSA does consider shorts to ground. This section provides             561 0-M-722A Rev. 0 [Attachment C section 5.0] - Nuclear Safety information related to the effects of a short to ground on typical ungrounded          Capability Fire Safe Shutdown Analysis Basis Document circuits.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                        Run: 06/26/2012 11:08        Page: 76 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.3 Circuit Failures Due to a Hot Short [A,  This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown equipment. A hot short is defined as a General]                                         fire-induced insulation breakdown between 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 Applicable Alignment Statement                              Alignment Basis                                                                        Reference NOT REQUIRED                                     This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                         Run: 06/26/2012 11:08  Page: 77 of 101 0                                                                                             0 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.3 Circuit Failures Due to a Hot Short [B,  A Hot Short on Grounded Circuits Grounded Circuits]
A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground as 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 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.
Applicability Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           The Turkey Point SSA does consider hot shorts. This section provides                  5610-M-722A Rev. 0 [Attachment C section 5.0] - Nuclear Safety information related to the effects of a hot short on typical grounded circuits.       Capability Fire Safe Shutdown Analysis Basis Document Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                       Run: 06/26/2012 11:08       Page: 78 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.                                   NEI 00-01 Guidance 3.5.2.3 Circuit Failures Due to a Hot Short [C,  A Hot Short on Ungrounded Circuits Ungrounded Circuits]
In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spurious operation. A single hot short can cause a spurious operation ifthe hot short comes from a circuit 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 Applicable Alignment Statement                              Alignment Basis                                                                       Reference Aligns                                           The Turkey Point SSA does consider hot shorts. This section provides                  561g-M-722A Rev. 0 [Attachment C section 5.0] - Nuclear Safety information related to the effects of a hot short on typical ungrounded circuits. Capability Fire Safe Shutdown Analysis Basis Document Turkey Point                                                      Run: 06/26/2012 11:08      Page: 79 of 101 0
Fire Safety Analysis Data Manager (4.1) 0                                                                                            is

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location.
Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be identified.
NEI 00-01 Ret.                                  NEI 00-01 Guidance 3.3.3.4 Identify Routing of Cables              Identify the routing for each cable including all raceway and cable endpoints. Typically, this information is obtained from joining the list of safe shutdown cables with an existing cable and raceway database Applicability Applicable Alignment Statement                             Alignment Basis                                                                    Reference Aligns                                          The routing for each required cable of a nuclear safety capability component is    561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe contained in the NFPA 805 database. The data was either imported from a            Shutdown Analysis (FSSA) controlled file developed during the validation project from a hard-copy plant     NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment document or provided via approved plant documentation. The information in          Database these sources have been verified to be accurate or were conservatively represented where direct verification of actual routing was not feasible.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                      Run: 06/26/2012 11:08         Page: 80 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NEI 00-01 Ref.                               NEI 00-01 Guidance 3.3.3.5 Identify Location of Raceway and     Identify the fire area location of each raceway and cable endpoint identified in the previous step and join this information with the cable routing data.
Cables by Fire Area                          In addition, identify the location of field-routed cable by fire area. This produces a database containing all of the cables requiring fire area analysis, their locations by fire area, and their raceway.
Applicability Applicable Alignment Statement                          Alignment Basis                                                                       Reference Aligns                                       The routing for each required cable of a nuclear safety capability component is       5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe contained in the NFPA 805 database. The data was either imported from a               Shutdown Analysis (FSSA) controlled file developed during the validation project from a hard-copy plant        NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment document or provided via approved plant documentation. The information in             Database these sources have been verified to be accurate or were conservatively represented where direct verification of actual routing was not feasible. The routing information includes the raceway to fire area correlation.
Run: 06/26/2012 11:08        Page: 81 of 101 0
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point 0                                                                                         0

00 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NEI 00-01 Ref.                                NEI 00-01 Guidance 3.5.2.4 Circuit Failures Due to Inadequate    The evaluation of associated circuits of a common power source consists of verifying proper coordination between the supply breaker/fuse and the Circuit Coordination                          load breakers/fuses for power sources that are required for safe shutdown. The concern is that, for fire damage to a single power cable, lack of 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 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 criteria in Appendix R, NRC staff guidance, and plant licensing bases.
Applicability Applicable Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                     Run: 06/26/2012 11:08       Page: 82 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location.
Alignment Statement                          Alignment Basis                                                                 Reference Aligns with Intent                           The circuit analysis procedure assumes coordination exists. In the SER dated    5610-M-722A Rev. 0 [Attachment C section 3.2] - Nuclear Safety April 16, 1984, the staff stated (based on the FPL submittal) that because      Capability Fire Safe Shutdown Analysis Basis Document coordination of circuit protective devices was part of the original electrical  NRC-FPL Letter 4/16/84 [Section Il.B.1] - Alternate Safe Shutdown system design, associated circuits by common power supply should not exist. Capability - Fire Protection, Sections III.G.3 and III.L of Appendix R However, this review was for the plant's alternate shutdown capability, and all to 10 CFR 50, Turkey Point Plant Units 3 and 4 power supplies being credited in the re-validation may not have been evaluated.
ITEM FOR IMPLEMENTATION (see PTN NFPA 805 LAR Attachment S Table S-2 Items 19 and 20):
* Perform modifications as necessary to achieve coordination on DP412A, 4DP86, and 4DP87 (only NSCA power supplies without demonstrated coordination)
This task will be performed by FPL.
Fire Safety Analysis Data Manager (4.1)                                                     Turkey Point                                                    Run: 06/26/2012 11:08         Page: 83 of 101 0                                                                                      0 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.3 Nuclear Safety Equipment and Cable Location.
NEI 00-01 Ref.                               NEI 00-01 Guidance 3.5.2.5 Circuit Failures Due to Common       The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to fire damage to a circuit either whose Enclosure Concerns                           isolation device fails to isolate the cable fault or protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.
The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.
Applicability Applicable Alignment Statement                          Alignment Basis                                                                         Reference Aligns                                       A common enclosure study of the plant's 4kV switchgears was performed and               5610-M-722A Rev. 0 [Attachment F] - Nuclear Safety Capability Fire documented in the Analysis Basis Document. All other power supplies were                Safe Shutdown Analysis Basis Document determined to not require external power for their overcurrent trip function.           STD-M-006 Rev. 1 [Section 5.4.2.3] - Turkey Point Unit 3 & 4 Fire Circuit protection is provided such that due to designed operation of the circuit       Protection Standard protective devices, circuits will not carry sufficient energy to cause cable damage. Cable size selection is based on not exceeding the cable design ratings prior to completion of protective action.
Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                         Run: 06/26/2012 11:08     Page: 84 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
Fire Area Assessment. An engineering analysis shall be performed in accordance with the 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 Assessment and Compliance             By determining the location of each component and cable by fire area and using the cable to equipment relationships described above, the affected Assessment                                          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 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 Applicable Alignment Statement                                 Alignment Basis                                                                      Reference NOT REQUIRED                                        This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
0 Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                       Run: 06/26/2012 11:08     Page: 85 of 101 0                                                                                               0

0                                                                                      0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                              NEI 00-01 Guidance 3.4.1 Criteria / Assumptions                The following criteria and assumptions apply when performing fire area compliance assessment to mitigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.
Applicability Applicable Alignment Statement                         Alignment Basis                                                                      Reference NOT REQUIRED                                This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                 Run: 06/26/2012 11:08      Page: 86 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                               NEI 00-01 Guidance 3.4.1.1 [Number of Postulated Fires]         Assume only one fire in any single fire area at a time.
Applicability Applicable Alignment Statement                          Alignment Basis                                                                   Reference Aligns                                       The analysis assumes the postulated fire is confined to one area, and that no     561 0-M-722A Rev. 0 [Attachment D sections 4.1 and 5.0] - Nuclear other fires occur. For the Outdoor Area, the analysis was done on a fire zone     Safety Capability Fire Safe Shutdown Analysis Basis Document basis, with the single fire affecting the fire zone and the adjacent outdoor fire zones within a "zone of influence". The zone of influence was 20 feet unless an approved exemption or the presence of significant combustibles allowed /
required the use of a different zone of influence.
Fire Safety Analysis Data Manager (4.1)                                                                                                                       Run: 06/26/2012 11:08 0
Page: 87 of 101 0                                                                                        Turkey Point 0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                              NEI 00-01 Guidance 3.4.1.2 [Damage to Unprotected Equipment    Assume that the fire may affect all unprotected cables and equipment within the fire area. This assumes that neither the fire size nor the fire intensity and Cables]                                 is known. This is conservative and bounds the exposure fire that is required by the regulation.
Applicability Applicable Alignment Statement                         Alignment Basis                                                                   Reference Aligns                                      The analysis assumes all unprotected cables and electrical and pneumatic          561 0-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety equipment can be affected and considers all potential failures in each area       Capability Fire Safe Shutdown Analysis Basis Document analyzed.                                                                         561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
Fire Safety Analysis Data Manager (4.1)                                                    Turkey Point                                                     Run: 06/26/2012 11:08         Page: 88 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                               NEI 00-01 Guidance 3.4.1.3 [Assess Impacts to Required          Address all cable and equipment impacts affecting the required safe shutdown path in the fire area. All potential impacts within the fire area must be Components]                                  addressed. The focus of this section is to determine 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.
Applicability Applicable Alignment Statement                          Alignment Basis                                                                   Reference Aligns                                       All potential impacts to the components and systems credited for safe             5610-M-722A Rev. 0 [Attachment D sections 4.0 and 5.0] - Nuclear shutdown in each fire area / analysis area are addressed in the analysis.         Safety Capability Fire Safe Shutdown Analysis Basis Document 561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA) is Fire Safety Analysis Data Manager (4.1)                                                     Turkey Point 0
Run: 06/26/2012 11:08      Page: 89 of 101 0

0                                                                                                                                                                             0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                              NEI 00-01 Guidance 3.4.1.4 [Manual Actions]                    Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in accordance with NRC requirements.
Applicability Applicable Alignment Statement                         Alignment Basis                                                                 Reference Aligns with Intent                          Manual actions are used throughout the safe shutdown analysis. All manual       561 O-M-722A Rev. 0 [Attachment D section 5.6.7] - Nuclear Safety actions are being evaluated for their risk significance to support safe         Capability Fire Safe Shutdown Analysis Basis Document shutdown. All risk significant actions have had their feasibility evaluated. 561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
Fire Safety Analysis Data Manager (4.1)                                                      Turkey Point                                                 Run: 06/26/2012 11:08      Page: 90 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment 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 Applicable Alignment Statement                        Alignment Basis                                                                Reference Aligns                                     The compliance assessment procedure requires that the specified criteria       5610-M-722A Rev. 0 [Attachment D section 5.6.7] - Nuclear Safety (equipment and tools available and staged, procedures, and adequate man        Capability Fire Safe Shutdown Analysis Basis Document power) be met in order to use a "repair" resolution code.
Fire Safety Analysis Data Manager (4.1) is Turkey Point 0
Run: 06/26/2012 11:08 is Page: 91 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                                NEI 00-01 Guidance 3.4.1.6 [Assess Compliance with Deterministic Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or Criteria]                                     emergency control station(s) is free of fire damage (III.G.1 .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 nonsafety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (llI.G.2.a)
                                              - Separation of cables and equipment and associated nonsafety 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 (lll.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 nonsafety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (lll.G.2.d);
                                              - Installation of fire detectors and an automatic fire suppression system in the fire area (lll.G.2.e); or
                                              - Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (lll.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 Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns                                        The deterministic criteria are part of the acceptable compliance strategies          561 0-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety used in the revalidaton.                                                             Capability Fire Safe Shutdown Analysis Basis Document NICAD Rev. 1.0a [Resolution Code Templates Table] - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                          Run: 06/26/2012 11:08  Page: 92 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                              NEI 00-01 Guidance 3.4.1.7 [Consider Additional Equipment]     Consider selecting other equipment that can perform the same safe shutdown function as the 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 Applicable Alignment Statement                         Alignment Basis                                                                 Reference Aligns                                      Part of the process used to perform fire area compliance assessment included    5610-M-722A Rev. 0 [Attachments B and D] - Nuclear Safety evaluating all paths to achieve a key safety function to determine which were   Capability Fire Safe Shutdown Analysis Basis Document available / least impacted by the postulated fire.
Fire Safety Analysis Data Manager (4.1)                                                     Turkey Point                                                   Run: 06/26/2012 11:08      Page: 93 of 101 0                                                                                       0 0

0 Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                               NEI 00-01 Guidance 3.4.1.8 [Consider Instrument Tubing Effects] Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent effects on instrument readings or signals associated with the protected safe shutdown path in evaluating post-fire safe shutdown capability. This can be done systematically or via procedures such as Emergency Operating Procedures.
Applicability Applicable Alignment Statement                          Alignment Basis                                                                       Reference Aligns                                       The potential effects of fire damage to instrument tubing has been considered.        5610-M-722A Rev. 0 [Section 3.15] - Nuclear Safety Capability Fire The only safe shutdown component with instrument tubing which crosses                 Safe Shutdown Analysis Basis Document analysis area boundaries are the atmospheric dump valves (ADVs - going                5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe from the steam platforms to the control room for control functions). The tubing       Shutdown Analysis (FSSA) was evaluated for control function, and to confirm that localized effects will not    PTN-FPER-10-009 Rev. 1 [RFI-0148] - SSA Upgrade Design adversely affect control signals or indications that would affect post-fire           Support Input shutdown capability.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                       Run: 06/26/2012 11:08      Page: 94 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                               NEI 00-01 Guidance 3.4.2 Methodology for Fire Area Assessment   Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area 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 Applicable Alignment Statement                          Alignment Basis                                                                       Reference NOT REQUIRED                                 This is an introductory paragraph with no specific criteria. Requirements are in the subsequent subsections.
Fire Safety Analysis Data Manager (4.1) is Turkey Point 0
Run: 06/26/2012 11:08 is Page: 95 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                                  NEI 00-01 Guidance 3.4.2.1 Identify the Affected Equipment by Fire Identify the safe shutdown cables, equipment and systems located in each fire area that may be potentially damaged by the fire. Provide this Area                                            information in a report format. The report may be 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).
Applicability Applicable Alignment Statement                             Alignment Basis                                                                   Reference Aligns                                          Both the existing analysis and the NFPA 805 Database provide reports of all       561 0-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety potentially damaged equipment in the fire area (or analysis area) being           Capability Fire Safe Shutdown Analysis Basis Document evaluated.                                                                        561 0-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe Shutdown Analysis (FSSA)
NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment Database Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                     Run: 06/26/2012 11:08      Page: 96 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment NEI 00-01 Ref.                                NEI 00-01 Guidance 3.4.2.2 Determine the Shutdown Paths Least    Based on a review of the systems, equipment and cables within each fire area, determine which shutdown paths are either unaffected or least Impacted By a Fire in Each Fire Area          impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipment in the 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.
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 Applicable Alignment Statement                           Alignment Basis                                                                      Reference Aligns                                        All systems and equipment within the analysis area are considered and a              5610-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety determination is made of which safe shutdown systems (forming a shutdown             Capability Fire Safe Shutdown Analysis Basis Document path) are least affected for performing the safe shutdown function. This             5610-M-722B Rev. 0 - Nuclear Safety Capability Fire Safe determination includes consideration of the availability of any supporting           Shutdown Analysis (FSSA) equipment or systems whose operation is necessary for a given component. These systems and their associated components, supporting systems, and cables are then evaluated for any fire effects which are then dispositioned.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                       Run: 06/26/2012 11:08      Page: 97 of 101 0                                                                                        0                                                                                            is

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area AssessmenL NEI 00-01 Ref.                               NEI 00-01 Guidance 3.4.2.3 Determine Safe Shutdown Equipment    Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine the equipment that can impact safe shutdown Impacts                                     and that can potentially be impacted by a fire in the fire area, and what those possible impacts are.
Applicability Applicable Alignment Statement                         Alignment Basis                                                                        Reference Aligns                                      The NFPA 805 database generates a list of all credited equipment that may be           561 0-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety impacted by a fire in each fire area / analysis area. The possible impacts to          Capability Fire Safe Shutdown Analysis Basis Document those components are then assessed based on that component's function(s)               NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment in that fire area and the function of any damaged cables.                              Database Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                       Run: 06/26/2012 11:08     Page: 98 of 10 1

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                                  NEI 00-01 Guidance 3.4.2.4 Develop a Compliance Strategy or        The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-2):
Disposition to Mitigate the Effects Due to Fire - Provide a qualified 3-fire rated barrier.
Damage to Each Required Component or Cable      - Provide a 1-hour fire rated barrier with automatic suppression and detection.
                                                - Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no intervening combustibles within the 20 foot separation distance.
                                                - Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability.
                                                - Provide a procedural action in accordance with regulatory requirements.
                                                - 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 IIl.G.2.d, e and f.
Applicability Applicable Alignment Statement                             Alignment Basis                                                                       Reference Aligns                                          The compliance strategies (resolution codes) are entered into the NFPA 805            5610-M-722A Rev. 0 [Attachment D Section 5.0] - Nuclear Safety database in accordance wtih the compliance assessment project instruction.            Capability Fire Safe Shutdown Analysis Basis Document Each cable associated with a credited component in each fire area is assigned a resolution code.
Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                        Run: 06/26/2012 11:08    Page: 99 of 101 0                                                                                         0                                                                                           0

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                                 NEI 00-01 Guidance 3.4.2.5 Document the Compliance Strategy or    Assign compliance strategy statements or codes to components or cables to identify the justification or mitigating actions proposed for achieving safe Disposition Determined to Mitigate the Effects shutdown. The justification should address the cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide each Due to Fire Damage to Each Required            piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or cable for Component or Cable                             the required safe shutdown path with a specific compliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area Assessment Report documenting each cable disposition.
Applicability Applicable Alignment Statement                            Alignment Basis                                                                    Reference Aligns                                         A compliance strategy code was applied to every combination of component          5610-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety and cable in the identified safe shutdown paths, including all components         Capability Fire Safe Shutdown Analysis Basis Document whose spurious operation or mal-operation could impact the ability to safely shut down. Once the potential fire effects and strategy codes on all credited equipment in the fire area / analysis area have been entered in the database, an area summary report is developed to document the overall status of the area.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                      Run: 06/26/2012 11:08       Page: 100 of 101

Attachment B NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review 2.4.2.4 Fire Area Assessment.
NEI 00-01 Ref.                               NEI 00-01 Guidance 3.5.1.5 [C, Likelihood of Undesired          Determination of the potential consequence of the damaged associated circuits is based on the examination of specific NPP piping and Consequences]                                instrumentation diagrams (P&IDs) and review of components 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 Applicable Alignment Statement                          Alignment Basis                                                                       Reference Aligns                                       During the component selection process, potential consequence of the                  561 0-M-722A Rev. 0 [Attachment D section 5.0] - Nuclear Safety damaged equipment (i.e., consequence will affect ability to achieve and               Capability Fire Safe Shutdown Analysis Basis Document maintain hot standby (HSD) or cold shutdown (CSD)) were identified. This              NICAD Rev. 1.0a - PTN NFPA 805 Compliance Assessment identification occurred through the use of P&IDs, Operating Procedures and            Database basis documents. The review considered components whose spurious operation or mal-operation could affect the shutdown path. During compliance assessment, the equipment's function as it pertains to the requirement to achieve and maintain HSD/CSD was considered and mitigative actions were binned based on those functions.
Fire Safety Analysis Data Manager (4.1)                                                       Turkey Point                                                        Run: 06/26/2012 11:08      Page: 101 of 101 is 0

Florida Power & Light                      Attachment E - Radioactive Release Transition Transition Florida Power & Liqht                      Attachment E Radioactive Release E. NEI 04-02 Radioactive Release Transition 23 Pages Attached Revision 0                                                                      Page E-1

Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Florida Power & Light                                                                              Attachment E Radioactive Release Transition Compartment           Auxiliary Building Compartment Selection and Justification Basis The Auxiliary Building is a shared building for both Unit 3 and Unit 4. This hardened safety related structure has consistent features to allow a common evaluation.
Auxiliary Building Fire Area(s)                                   Auxiliary Building Pre-Fire Plan(s)                    Rev.
A          Unit 3 and 4 Auxiliary Building                         001     Units 3 and 4 Holdup Tank A                                02/09/01 AAA        Unit 3 and 4 Auxiliary Building                         002     Units 3 and 4 Holdup Tank B                                02/09/01 B         Unit 3 RHR Area                                         003     Units 3 and 4 Holdup Tank C                                02/09/01 BBB       Unit 3 and 4 Safety Injection Pump Rooms                004     Unit 3 and 4 Hallway South of Holdup Tanks                 02/09/01 C         Unit 4 RHR Area                                         005     Units 3 and 4 Chemical Drain, Laundry and Hot Shower       08/12/02 D         Unit 4 Pipe and Valve Room                                     Tank Room E         Unit 3 Pipe and Valve Room                              006     Units 3 and 4 Gas Compressor Room                          08/12/02 F         Unit 3 and 4 Auxiliary Building                         007     Units 3 and 4 Gas Compressor Room                          08/12/02 G         Units 3 and 4 Electrical Equipment Room                 008     Units 3 and 4 Waste Holdup Tank Room                       08/12/02 H         Unit 3 West Electrical Penetration Room                 009     Units 3 and 4 Waste Evaporator Feed Pump Room              08/12/02 I         Unit 3 South Electrical Penetration Room                010     Units 3 and 4 Radioactive Pipeway                            REV 1 J         Unit 4 North Electrical Penetration Room                011     Unit 3 RHR Heat Exchanger Room                              12/21/09 K         Unit 4 West Electrical Penetration Room                 012     Unit 3 RHR Pump A Room                                     06/05/08 L         Units 3 and 4 Auxiliary Building Fan Room               013     Unit 3 RHR Pump B Room                                     08/16/04 N         Unit 4 Charging Pump Room                               014     Unit 4 RHR Heat Exchanger Room                              12/21/09 0         Unit 3 Charging Pump Room                               015     Unit 4 RHR Pump A Room                                     08/12/02 016     Unit 4 RHR Pump B Room                                     08/16/04 017     Units 3 and 4 Spent Resin Storage Tank Room                02/09/01 018     Units 3 and 4 Area Beneath Boric Acid Evaporator Skid      02/09/01 019     Unit 3 West Electrical Penetration Room                    05/21/08 020     Unit 3 South Electrical Penetration Room                   02/09/01 021     Units 3 and 4 Counting Room                                02/09/01 022     Units 3 and 4 Hot Lab                                      09/13/06 023     Units 3 and 4 Gas Decay Tank                               02/09/01 024     Units 3 and 4 Gas Decay Tank Room                          02/09/01 025     Units 3 and 4 Electrical Equipment Room                      REV 1 025A Units. 3 and 4 Spare Battery Room                             02/09/01 Revision 0                                                                                                                                 Page E-2

Florida Power & Light                                            Attachment E - Radioactive Release Transition Florida Power & Light                                            Attachment E Radioactive Release Transition Compartment       Auxiliary Building 026 Unit 4 North Electrical Penetration Room                 08/12/02 027 Unit 4 West Electrical Penetration Room                  08/12/02 028 Units 3 and 4 Auxiliary Building Fan Room                10/23/09 029 Unit 4 Spent Fuel Pit                                    08/12/02 030 Unit 4 Pipe and Valve Room                               09/23/04 031 Unit 4 Containment Spray Pump Room                       08/12/02 032 Unit 4 Sample Room                                       08/12/02 033 Units 3 and 4 Post Accident Sampling System Room         08/12/02 034 Unit 4 Boric Acid Evaporator Package Room                08/12/02 035 Units 3 and 4 Valve Room                                 08/12/02 036 Unit 3 Boric Acid Evaporator Package Room                08/12/02 037 Unit 3 Sample Room and Gas Stripper Room                 08/12/02 038 Unit 3 Containment Spray Pump Room                       08/12/02 039 Unit 3 Concentrate Holding Tank Room                     08/12/02 040 Unit 3 Pipe and Valve Room                               09/23/04 041 Units 3 and 4 Boric Acid Tanks and Pump Room             02/09/01 042 Unit 3 Spent Fuel Pit                                    08/12/02 043 Unit 4 Spent Fuel Pit Heat Exchanger Room                08/12/02 044 Unit 4 New Fuel Storage                                  08/12/02 045 Unit 4 Charging Pump Room                                08/12/02 046 Units 3 and 4 Laundry Room                               08/12/02 048 Units 3 and 4 Deborating Demineralizer Tank Room         02/09/01 049 Units 3 and 4 Base and Cation Radwaste                   08/12/02 Demineralizers Room 050 Units 4 Purification Demineralizers Room                 08/12/02 051 Units 3 and 4 Condensate Pump and Monitor Tank Room      08/12/02 052 Unit 4 Safety Injection Pump Room                         REV 1 053 Unit 3 Safety Injection Pump Room                         REV 1 055 Unit 3 Charging Pump Room                                02/09/01 056 Unit 3 New Fuel Storage Room                             08/12/02 057 Unit 3 Spent Fuel Pit Heat Exchanger Room                08/12/02 058 Units 3 and 4 Auxiliary Building Hallway-Elev. 18'        REV 1 Revision 0                                                                                            Page E-3

Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Compartment             Auxiliary Building 111     Unit 4 Volume Control Tank Room                              12/20/06 112     Unit 3 Volume Control Tank Room                              12/20/06 Smoke and By Products of Combustion- Airborne Effluent Evaluation The Auxiliary Building Ventilation System is designed to maintain a negative pressure throughout the building which will allow air to enter the building, but effluent not to escape. Smoke would be maintained within the building. Exhaust air from the Auxiliary Building Ventilation System is routed to atmosphere via the plant vent. Monitors are in continuous service in the plant vent. Unit 3 also includes the Spent Fuel Pool stack vent. This flow path is also continuously monitored. Plant operators will take action ifa high activity alarm is received in the control room.
Fire Suppressant Run Off- Liquid Effluent Evaluation Floor drains located within the Auxiliary Building drain to the waste hold up tank. This tank is capable of holding 24,300 gallons of waste liquid.
The waste hold up tank within the Auxiliary Building can then be pumped to the Radwaste Building for processing.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations Ventilation and drainage engineering controls are not affected by the unit operating condition. However, the likelihood of doors and hatches being open during outage time periods is much greater. Administrative guidance for the fire brigade will drive actions to prevent uncontrolled radioactive release in this scenario.
Conclusion Using installed engineering controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Revision 0                                                                                                                                   Page E-4

Florida Power & Light                                                                                Attachment E - Radioactive Release Transition Compartment             Dress Out Building Compartment Selection and Justification Basis The Dress Out Building houses protective clothing and other radiation protection equipment for use in the RCA. This is a stand alone structure within the RCA. This building is separate from the power block and has individual features and methods of compliance.
Dress Out Building Fire Area(s)                                  Dress Out Building Pre-Fire Plan(s)                     Rev.
OD          Outdoors                                                156     Dressout Facility                                          08/12/02 Smoke and By Products of Combustion- Airborne Effluent Evaluation The Dress Out Building contains no specific ventilation system used to maintain or control contamination within the building. Administrative control will be used to maintain compliance in this area.
Fire Suppressant Run Off- Liquid Effluent Evaluation No specific collection or drainage is identified in this area. Administrative controls will be used to maintain compliance with NFPA 805.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This facility is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are not impacted by the power condition.
Conclusion Revision 0                                                                                                                                  Page E-5

Florida Power& Light                                                                           Attachment E - Radioactive Release Transition Compartment             Dress Out Building Using administrative controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-6 Revision 0 Revision   0                                                                                                                           Page E-6

0                                                                                                                                                          4 Florida Power & Light                                                                                  Attachment E - Radioactive Release Transition Compartment            Dry Storage Building Compartment Selection and Justification Basis The Dry Storage Building is used to store contaminated equipment and materials. This is a stand alone structure within the RCA. This building is separate from the power block and has individual features and methods of compliance.
Dry Storage Building Fire Area(s)                                    Dry Storage Building Pre-Fire Plan(s)                     Rev.
OD         Outdoors                                                 157     Dry Storage Building                                           08/12/02 Smoke and By Products of Combustion- Airborne Effluent Evaluation The Dry Storage Building is equipped with exhaust ventilation as well as a ridge vent along the roof line of the building. No exhaust monitors or controls are installed within this building. The building is equipped with a sprinkler system. This system will act as smoke scrubbing in the event of a major fire. This will not only assist in minimizing the size of the fire, but again, driving particulate matter out of the smoke and down to the ground.
Fire Suppressant Run Off- Liquid Effluent Evaluation The building has a berm created by the foundation. The berm is in place on all sides of the building. Floor drains are routed to a building sump. The building sump is sampled prior to release.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This facility is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are Revision 0                                                                                                                                      Page E-7

Florida Power & Light                                                                             Attachment E - Radioactive Release Transition Compartment           Dry Storage Building not impacted by the power condition.
Conclusion Using installed engineered and administrative controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Revision 0                                                                                                                            Page E-8 0

Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Florida Power &Light                                                                               Attachment E Radioactive Release Transition Compartment            Non-RCA Compartment Selection and Justification Basis The Non-RCA areas compartment consists of those fire area/fire zones that are not part of the radiologically controlled area of the unit. The individual Fire Areas, Fire Zones and pre-fire plans may be contained in other RCA compartments due to a portion of the fire area, zone and/or pre-fire plan being within the RCA. The non-RCA portions of the fire areas and pre-fire plans listed do not store or contain radioactive material that could result in the potential for a radioactive release. The areas not within the RCA screen out of the evaluation process.
Non-RCA Fire Area(s)                                             Non-RCA Pre-Fire Plan(s)                          Rev.
AA         Unit 3 Train B Emergency Diesel Generator Day           061    Unit 4 Reactor Control Rod Equipment Room                    REV 1 Tank Rm                                                 062    Units 3 and 4 Computer Room                                 11/08/07 AAA        Unit 3 and 4 Auxiliary Building                         063    Unit 3 Reactor Control Rod Equipment Room                    REV 1 BB         Unit 3 Train A Emergency Diesel Generator Day           064    Units 3 and 4 Vestibule Elevator                           08/12/02 Tank Rm                                                 065    Units 3 and 4 QSPDS Inverter Area                          02/09/01 CCC        Units 3 and 4 Vestibule Elevator                        066    Unit 4 Steam Generator Feed Pump Area                      02/09/01 DD         Unit 4 480V Load Centers A and B Room                   067    4160V Switchgear 4B Room                                     REV 2 DDD        Unit 4 Steam Generator Feed Pump Area                   068    Unit 4 4160 V Switchgear 4A Room                             REV 2 EE         Unit 4 480V Load Centers C and D Room                   069    Unit 3 Steam Generator Feed Pump Area                      02/09/01 EEE        Unit 3 Steam Generator Feed Pump Area                   070    4160V Switchgear 3B Room                                     REV 1 FF         Unit 3 480V Load Centers A and B Room                   071    4160V Switchgear 3A Room                                     REV 1 FFF        Units 3 and 4 Purge Supply Fan Room                     072    Unit 3B Emergency Diesel Generator Building                  REV 1 GG         Unit 3 480V Load Centers C and D Room                   073    Unit 3A Emergency Diesel Generator Building                06/02/09 HH         Units 3 and 4 Cable Spreading & Chase                   074    Unit 3B Emergency Diesel Generator Day Tank Room           08/12/02 II         Unit 4 B DC Equipment Room                              075    Unit 3A Emergency Diesel Generator Day Tank Room           06/02/09 JJ         Unit 4 Battery Rack B Room                              076    Unit 4 Lube Oil Reservoir Area                             02/28/05 KK         Unit 3 Battery Rack A Room                              077    Unit 4 Laydown Area, Instrument Air Compressors and        08/12/02 LL         Unit 3 A DC Equipment Room                                     Condensate Storage Area MM         Units 3 and 4 Control Room                              078    Unit 4 Instrument Air Equipment Area                       02/09/10 NN         Units 3 and 4 A DC Equipment Room                       079    Outdoor Area West of Unit 4 Containment                      REV 1 OD         Outdoors                                                080    Unit 4 Main Condenser Area                                 08/12/02 00         Units 3 and 4 B DC Equipment Room                       081    Unit 4 Main and Start-up Transformer and Unit 3 Turbine     12/21/09 PP         Unit 4 Battery Rack A Room                                     Lube Oil Reservoir Area QQ         Unit 3 Battery Rack B Room                              082    Unit 4 Auxiliary Transformer Area                          06/03/09 R          Unit 4 Reactor Control Rod Equipment Room               083    Unit 3 Instrument Air Equipment Area                       02/09/10 Page E-9 Revision 00                                                                                                                                Page E-9

Florida Power & Light                                                                      Attachment E - Radioactive Release Transition Florida Power & Light                                                                     Attachment E Radioactive Release Transition Compartment          Non-RCA RR        Unit 4 Train A Emergency Diesel Generator Room     084  Unit 3 and 4 Auxiliary Feed Pump Area                     REV 1 S         Units 3 and 4 Computer Room                        085  Unit 3 Main Condenser Area                               08/12/02 SS        Unit 4 Train B Emergency Diesel Generator Room     086  Unit 3 Main Transformer and Startup Transformer Area     08/12/02 T         Unit 3 Reactor Control Rod Equipment Room          087  Unit 3 Auxiliary Transformer Area                        06/03/09 TT        Unit 3 Switchgear Room 3D                          088  Unit 3 Switchgear/Diesel Generator Building Vestibule    08/12/02 U         4160 V Switchgear 4B Room                          089  Unit 3 Condensate Storage Tank Area                      02/09/01 UU        Unit 4 Switchgear Room 4D                          090  Unit 3 Emergency Diesel Generator Oil Storage Tank       11/14/05 V         4160 V Switchgear 4A Room                          091  Unit 4 Condensate Pump Area                              08/12/02 VV        Unit 4 Train A Emergency Diesel Generator Control  092  Unit 3 Condensate Pump Area                              08/12/02 Rm.                                                093  Unit 4 480V Load Centers A and B Room                    08/16/04 W         4160 V Switchgear 3B Room                          094  Unit 4 480V Load Centers C and D Room                     REV 1 WW        Unit 4 Train A Diesel Oil Storage Tank &Pumps      095  Unit 3 480V Load Centers A and B Room                    02/09/01 X         4160 V Switchgear 3A Room                          096  Unit 3 480V Load Centers C and D Room                    02/09/01 XX        Unit 4 Train B Diesel Oil Storage Tank & Pumps     097  Units 3 and 4 Mechanical Equipment Room                   REV 1 Y         Unit 3 Train B Emergency Diesel Generator Building 098  Units 3 and 4 Cable Spreading Room                        REV 1 YY        Unit 4 Train B Emergency Diesel Generator Control  101  Unit 4 B DC Equipment Room                                REV 1 Rm.                                                102  Unit 4 Battery Rack B Room                               04/23/09 Z         Unit 3 Train A Emergency Diesel Generator Building 103  Unit 3 Battery Rack A Room                               12/03/09 104  Unit 3 A DC Equipment Room                               12/03/09 105  Units 3 and 4 Turbine Building Mezzanine Deck            12/03/09 106  Units 3 and 4 Control Room                                REV 1 106R Units 3 and 4 Control Room Roof                          08/12/02 107  Units 3 and 4 Purge Supply Fan Room                      12/03/09 108A Units 3 and 4 DC Equipment 4A Room                        REV 2 108B Units 3 and 4 DC Equipment 3B Room                        REV 1 109  Unit 4 Battery Rack A Room                               12/03/09 110  Unit 3 Battery Rack B Room                               04/23/09 113  Unit 4 Feedwater Platform                                08/12/02 114  Unit 4 Main Steam Header Platform                        08/12/02 115  Unit 3 Main Steam Header Platform                        08/12/02 116  Unit 3 Feedwater Platform                                08/12/02 117  Unit 3 and 4 Turbine Deck Area                           12/03/09 Revision 0                                                                                                                    Page E-10

Florida Power & Light                                 Attachment E - Radioactive Release Transition Florida Power & Light                                 Attachment E Radioactive Release Transition Compartment       Non-RCA 118 Units 3 & 4 Auxiliary Building Roof                      12/03/09 119 Unit 4 Circulating Water Intake Structure                02/09/10 120 Unit 3 Circulating Water Intake Structure                02/09/10 121 Unit 3 and 4 Intake Laydown Area                         02/09/10 122 Units 3 & 4 Water Treatment Area                         05/21/08 127 Units 1 & 2 Fuel Oil Storage Tanks                       08/12/02 128 Units 3 & 4 Distribution Switchyard                      08/12/02 129 Units 3 & 4 Nuclear Entrance Building                    05/21/08 131 Units 3A & 3B Diesel Generator Radiator Rooms            08/12/02 132 Units 3 & 4 Control Room Electrical Cable Chase          08/12/02 133 Unit 4B Emergency Diesel Generator Room                  02/09/01 134 Unit 3 Switchgear Room 3D                                09/04/07 135 Unit 4B Emergency Diesel Generator Control Room          02/09/01 136 Unit 4B Diesel Oil Transfer Pump Room                    02/09/01 137 Unit 4B Diesel Oil Storage Tank                          02/09/01 138 Unit 4A Emergency Diesel Generator Room                  02/09/01 139 Unit 4 Switchgear Room 4D                                09/04/07 140 Unit 4A Emergency Diesel Generator Control Room          02/09/01 141 Unit 4A Diesel Oil Transfer Pump Room                    02/09/01 142 Unit 4A Diesel Oil Storage Tank                          02/09/01 143 Unit 3 Emergency Diesel Generator Roof                   08/12/02 150 Area South of Units 3 and 4 Power Block                  05/21/08 151 Unit 3 Condensate Polishing Area                         05/30/03 152 Unit 4 Condensate Polishing Area                         05/21/08 153 Health-Physics Building                                  09/13/06 154 Nuclear Maintenance Building                             05/21/08 155 Nuclear Administration Building                          05/21/08 158 Central Receiving Building                               09/13/06 159 Waste Storage Building                                   08/12/02 160 Machine Shop Building                                    05/30/03 161 Issue Warehouse                                          5/21/08 162 Overflow Building                                        05/30/03 Revision 0                                                                                Page E-11

Florida Power & Light                                                                  Attachment E - Radioactive Release Transition Compartment         Non-RCA 163    Simulator Training Building                             08/16/04 164    Nuclear Maintenance Facility                             5/21/08 165    Diesel Generator Building/CAS                            4/28/05 166    Child Development Center                                02/09/01 167    Atrium Restaurant                                       02/09/01 999    Miscellaneous Area                                      02/09/01 Smoke and By Products of Combustion- Airborne Effluent Evaluation Non-RCA - Screens out of Radioactive Release Evaluation.
Fire Suppressant Run Off- Liquid Effluent Evaluation Non-RCA - Screens out of Radioactive Release Evaluation.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release Non-RCA - Screens out of Radioactive Release Evaluation.
Fire Brigade Training to Minimize Radioactive Release Non-RCA - Screens out of Radioactive Release Evaluation.
Non-Power Operations Non-RCA - Screens out of Radioactive Release Evaluation.
Conclusion Non-RCA - Screens out of Radioactive Release Evaluation.
Revision 0                                                                                                                 Page E-12

0                                                                          0 Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Compartment            Radwaste Building Compartment Selection and Justification Basis The Radwaste Building is a standalone structure. The hardened structure provides a natural means to contain smoke and fire suppression water runoff. The drainage and ventilation systems are specific to the building.
Radwaste Building Fire Area(s)                                   Radwaste Building Pre-Fire Plan(s)                      Rev.
AAA         Unit 3 and 4 Auxiliary Building                        126     Units 3 &4 Radwaste Building                                08/12/02 Smoke and By Products of Combustion- Airborne Effluent Evaluation The Radwaste Building exhaust ventilation is aligned to cause an in-flow of air from outside of the building. Exhaust air is discharged to atmosphere via the plant vent. Monitors are in continuous service in the area and actuate a high activity alarm in the control room if unacceptable levels are reached. Operators can take manual actions to stop a release on a high activity alarm.
Fire Suppressant Run Off- Liquid Effluent Evaluation Radwaste Building floor drains will drain to the waste hold up tank within the Radwaste Building. This 10,000 gallon tank is part of the radwaste system. This system will monitor and process waste until an acceptable condition is met to release or return waste water to a monitor tank for continued processing within the radwaste system.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This facility is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are Revision 0                                                                                                                                 Page E-13

Florida Power & Light                                                                          Attachment E - Radioactive Release Transition Compartment             Radwaste Building not impacted by the power condition.
Conclusion Using installed engineering controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-14 Revision 00 Revision                                                                                                                              Page E-14

0                                                                                                                                                     1 4
Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Compartment            RCA Access Facility Compartment Selection and Justification Basis This stand alone structure is the interface from RCA to non-RCA. This building is separate from the power block and has individual features and methods of compliance.
RCA Access Facility Fire Area(s)                                RCA Access Facility Pre-Fire Plan(s)                     Rev.
OD          Outdoors                                                168    RCA Access Facility                                            N/A NEW New Pre-Fire Plan is Needed                                       N/A Smoke and By Products of Combustion- Airborne Effluent Evaluation The RCA Access Facility is an interface location between RCA and non-RCA. The building houses the control point, source storage room, count room, calibration laboratory and other radiation protection support equipment. The building is equipped with an automatic sprinkler system that will provide some level of smoke scrubbing as well as containing a fire and reducing the overall amount of smoke created by the fire. Administrative controls will be used to ensure compliance with NFPA 805 Radioactive Release Criteria.
Fire Suppressant Run Off- Liquid Effluent Evaluation No special features to control water runoff were identified in the RCA Access Facility. Administrative controls will be used to ensure that potentially contaminated water runoff is contained.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This facility is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are Revision 0                                                                                                                                 Page E-15

Florida Power & Light                                                                          Attachment E - Radioactive Release Transition Compartment             RCA Access Facility not impacted by the power condition.
Conclusion Using administrative controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-16 Revision 00                                                                                                                           Page E-16

Florida Power & Light                                                                             Attachment E - Radioactive Release Transition Florida Power &Light                                                                              Attachment E Radioactive Release Transition Compartment            RCA-Outdoors/Yard Compartment Selection and Justification Basis The compartment RCA-Outdoors/Yard are locations open to atmosphere with no special drainage features or additional suppression systems to aid in preventing a radioactive release. This includes small structures, land/sea containers, and drums stored in locations where they have no engineering controls to prevent radioactive release.
RCA-Outdoors/Yard Fire Area(s)                                   RCA-Outdoors/Yard Pre-Fire Plan(s)                      Rev.
BBB        Unit 3 and 4 Safety Injection Pump Rooms               047     Unit 4 Component Cooling Pump and Heat Exchanger           03/05/10 CC         Units 3 and 4 Auxiliary Building North-South                   Area Breezeway                                              054     Unit 3 Component Cooling Pump and Heat Exchanger           08/12/02 OD         Outdoors                                                       Area 077     Unit 4 Laydown Area, Instrument Air Compressors and       08/12/02 Condensate Storage Area 079     Outdoor Area West of Unit 4 Containment                     REV 1 079A Units 3 and 4 Auxiliary Building North-South Breezeway       02/09/01 084 Unit 3 and 4 Auxiliary Feed Pump Area                           REV 1 118 Units 3 &4 Auxiliary Building Roof                             12/03/09 121     Unit 3 and 4 Intake Laydown Area                          02/09/10 122 Units 3 &4 Water Treatment Area                               05/21/08 123 Unit 3 & 4 Refueling Water Storage Tanks                      08/12/02 124     Outside Area NE of Unit 3 Containment                     08/12/02 125     Unit 3 and 4 Land Area South of Unit 4 Containment        08/12/02 150 Area South of Units 3 and 4 Power Block                       05/21/08 999     Miscellaneous Area                                        02/09/01 Smoke and By Products of Combustion- Airborne Effluent Evaluation The yard area is open to atmosphere and has no ventilation system to assist or act as an engineering control. Administrative controls will be used to ensure compliance with NFPA 805.
Fire Suppressant Run Off- Liquid Effluent Evaluation The yard area is equipped with storm drains that will discharge into the canal system. Administrative controls will be used to ensure release to the canal system is controlled and monitored.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release Revision 0                                                                                                                              Page E-17

Florida Power & Light                                                                               Attachment E - Radioactive Release Transition Compartment             RCA-Outdoors/Yard The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This area is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are not impacted by the power condition.
Conclusion Using administrative controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-18 Revision 00 Revision                                                                                                                                   Page E-18

Florida Power & Light                                                                               Attachment E - Radioactive Release Transition Compartment             Steam Generator Storage Compound Compartment Selection and Justification Basis This structure is similar to others in the yard area. Ithas been separated to call out the need for a specific pre-fire plan to be developed for this building. The adjacent RCP Maintenance Building and Storage Building should be included in this Pre-Fire Plan.
Steam Generator Storage Compound Fire Area(s)                     Steam Generator Storage Compound Pre-Fire Plan(s)                   Rev.
OD          Outdoors                                               NEW New Pre-Fire Plan is Needed                                        N/A Smoke and By Products of Combustion- Airborne Effluent Evaluation No specific ventilation system is installed in the Storage Building. Administrative controls will be used to maintain compliance with NFPA 805.
Fire Suppressant Run Off- Liquid Effluent Evaluation The building is constructed of concrete with a 6-inch high water stop integrated into the foundation and will maintain any liquids for fire suppression within the area.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations This facility is not impacted by plant operating condition or power levels. All compliance strategies to meet NFPA 805 Radioactive Release are not impacted by the power condition.
Conclusion Revision 0                                                                                                                                 Page E-19

Florida Power & Light                                                                          Attachment E - Radioactive Release Transition Florida Power & Light                                                                          Attachment E Radioactive Release Transition Compartment             Steam Generator Storage Compound Using administrative controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Revision 0                                                                                                                            Page E-20 0

Florida Power & Light                                                                               Attachment E - Radioactive Release Transition Compartment             Unit 3 Containment Compartment Selection and Justification Basis The containment building is a hardened structure designed to hold in sources of radioactive contamination. The engineering controls are specific to the containment building.
Unit 3 Containment Fire Area(s)                                  Unit 3 Containment Pre-Fire Plan(s)                     Rev.
Q          Unit 3 Containment Building                             060     Unit 3 Containment Building                                  REV 1 Smoke and By Products of Combustion- Airborne Effluent Evaluation A normal containment cooling system is in place to cool and recirculate air through the containment. There is no location for air to escape or enter under normal power conditions. The containment purge system allows for the removal of any air within the containment to go through particulate filters and monitoring prior to release. Containment purge will release through the plant vent which is continuously monitored.
Fire Suppressant Run Off- Liquid Effluent Evaluation Any water used for fire suppression activities within containment would drain to the containment sump. The containment sump can then be pumped into the radwaste system for processing and releasing in accordance with station procedures and technical specifications.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations During non-power operations, such as refueling outages, hatches may be opened and airlocks may be bypassed. During these time periods, administrative controls will maintain compliance with NFPA 805 radioactive release criteria.
Revision 0                                                                                                                                 Page E-21

Florida Power & Light                                                                          Attachment E - Radioactive Release Transition Florida Power & Light                                                                          Affachment E Radioactive Release Transition Compartment             Unit 3 Containment Conclusion Using installed engineering controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-22 Revision 00                                                                                                                            Page E-22

0                                                                          0 Florida Power & Light                                                                              Attachment E - Radioactive Release Transition Compartment             Unit 4 Containment Compartment Selection and Justification Basis The containment building is a hardened structure designed to hold in sources of radioactive contamination. The engineering controls are specific to the containment building.
Unit 4 Containment Fire Area(s)                                  Unit 4 Containment Pre-Fire Plan(s)                     Rev.
P          Unit 4 Containment Building                             059     Unit 4 Containment Building                                 08/12/02 Smoke and By Products of Combustion- Airborne Effluent Evaluation A normal containment cooling system is in place to cool and recirculate air through the containment. There is no location for air to escape or enter under normal power conditions. The containment purge system allows for the removal of any air within the containment to go through particulate filters and monitoring prior to release. Containment purge will release through the plant vent which is continuously monitored.
Fire Suppressant Run Off- Liquid Effluent Evaluation Any water used for fire suppression activities within containment would drain to the containment sump. The containment sump can then be pumped into the RW system for processing and releasing in accordance with station procedures and technical specifications.
Administrative Controls - Pre-Plans, Procedures, and Guidelines to Minimize the Risk of Radioactive Release The site pre-fire plans will be modified to include references and recommendations for minimizing the likelihood of a radioactive release. Fire Brigade standard operating procedures will be developed to provide guidance and expectations for response to fires involving radioactive material.
The radiation protection process for support of the fire brigade will be incorporated into fire response standard operating procedures.
Fire Brigade Training to Minimize Radioactive Release The fire brigade training program will be modified to include objectives for identifying radioactive release paths and actions that can be taken by the fire brigade to prevent radioactive release.
Non-Power Operations During non-power operations, such as refueling outages, hatches may be opened and airlocks may be bypassed. During these time periods, administrative controls will maintain compliance with NFPA 805 radioactive release criteria.
Revision 0                                                                                                                                 Page E-23

Florida Power & Light                                                                          Attachment E - Radioactive Release Transition Florida Power & Light                                                                          Attachment E Radioactive Release Transition Compartment             Unit 4 Containment Conclusion Using installed engineering controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition.
Page E-24 Revision 00                                                                                                                           Page E-24

Florida Power & Liqht                      Attachment F - Fire-induced IVISOs Resolution F. Fire-Induced Multiple Spurious Operations Resolution 5 Pages Attached Revision 0                                                                      Page F-1

Florida Power & Light                             Attachment F - Fire-induced MSOs Resolution MSO Process Summary The following provides the guidance from FAQ 07-0038, Revision 3, along with the process and results.
Step I - Identify potential MSOs of concern Information sources that may be used as input include:
    " Post-fire safe shutdown analysis (NEI 00-01, Revision 1, Chapter 3)
    " Generic lists of MSOs (e.g., from Owners Groups and/or later versions of NEI 00-01, if endorsed by NRC for use in assessing MSOs)
    " Self assessment results (e.g., NEI 04-06 assessments performed to addressed RIS 2004-03)
    " PRA insights (e.g., NEI 00-01 Revision 1, Appendix F)
    " Operating Experience (e.g., licensee event reports, NRC Inspection Findings, etc.)
Results of Step 1:
The following information sources were used to identify the potential PTN MSOs of concern:
     " Post-fire safe shutdown analysis
     " PWROG generic MSO list
     " Fire PRA work completed at the time of expert panels
     " Internal events PRA insights Following the initial expert panel of 2008, an additional follow on review was performed in 2010 of the updated PWROG generic MSO list associated with NEI 00-01, Revision 2.
Step 2- Conduct an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Revision 1 Section F.4.2).
The expert panel should focus on system and component interactions that could impact nuclear safety. This information will be used in later tasks to identify cables and potential locations where vulnerabilities could exist.
The documentation of the results of the expert panel should include how the expert panel was conducted including the members of the expert panel, their experience, education, and areas of expertise. The documentation should include the list of MSOs reviewed as well as the source for each MSO. This documentation should provide a list the MSOs that were included in the PRA and a separate list of MSOs that were not kept for further analysis (and the reasons for rejecting these MSOs for further analysis).
Describe the expert panel process (e.g., when it was held, what training was provided to the panel members, what analyses were reviewed to identify MSOs, how was consensus achieved on which MSOs to keep and any dispute resolution process criteria used in decision process, etc.).
Page F-2 Revision 00                                                                          Page F-2

Florida Power & Light                             Attachment F - Fire-Induced MSOs Resolution
[Note: The physical location of the cables of concern (e.g., fire zone/area routing of the identified MSO cables), if known, may be used at this step in the process to focus the scope of the detailed review in further steps.
Results of Step 2:
Initial Expert Panel 2008 The initial expert panel consisted of a day and a half meeting at the PTN site with representatives from PTN fire protection and post-fire safe shutdown, PTN Operations/PRA, and supporting contract staff. The panel conducted document reviews, and held discussions on potential fire-induced spurious operations that could potentially impact plant safety. Documents that were used as guidance included:
     " Pressurized Water Reactor Owners Group (PWROG) Draft Working MSO list [file named PWR+MSO+List+rev+E.xls(file dated 3-25-08) from the NEI MSO Webboard]
     " NFPA 805 FAQ 07-0038, Lessons Learned on Multiple Spurious Operations, Revision 1 Training for the initial 2008 expert panel was conducted in the form of an introductory overview. Topics discussed included:
     " Purpose and scope of the safe shutdown
     " PRA overview
     " Overview training on the MSO issue, including o Background on Fire-Induced MSOs o Types of circuit failures that can occur and result in spurious operations o Results of the Fire Testing (EPRI/NEI Testing) o Role of the MSO resolution in NFPA 805 Transition Key points of the training included:
     " The proposed scenarios should not have presupposed limits on the number of fire-induced hot shorts or spurious operations (e.g., do not assume only one or two, one at a time, etc.).
     " The focus should not be on individual fire area locations, but rather on a system/component approach, in order to allow the analysis following the expert panel (e.g., PRA model and scenario development) to determine the vulnerability of the proposed interactions to credible fires.
A pre-job brief was conducted by a conference call in August 2008 with most of the expert panel participants to address the topics described above. The topics were reviewed again at the beginning of the onsite expert panel meeting.
The first day of the expert panel focused on a review of the PWROG scenarios. The PWROG Generic MSO list includes scenarios related to the following functions:
* Reactivity Control
     " RCS Inventory Control (Makeup)
     " RCS Pressure Control Revision 0                                                                           Page F-3

Florida Power & Light                           Attachment F - Fire-induced MSOs Resolution
    "   Decay Heat Removal
* Support Functions 0
By using the PWROG Generic MSO list as guidance, a step-by-step discussion was held, typically by reviewing P&IDs, postulating scenarios, discussing the potential consequences and likelihood, discussing operator response, and recommending additional courses of action. Key considerations, in addition to consequences were:
    " Whether the scenario of concern was currently modeled in the PTN SSA
    " Whether the scenario of concern was currently modeled in the PTN Internal Events PRA
    " Whether procedures addressed the potential scenarios of concern
    " Additional analyses or justification that may be necessary to document exclusion of a particular scenario
    " Potential generic issues that may be best resolved by PWROG analysis/justification In addition to the item-by-item review of the PWROG Generic MSO list, an additional "brainstorming" review was conducted by the Expert Panel in order to look for additional plant specific scenarios and general system pinchpoints that may not have existed specifically on the generic list. These reviews considered system flowpaths and addressed items such as deadheading of pumps, pump runout, and flow diversion. The additional reviews were conducted on the following key systems:
    " Auxiliary Feedwater (AFW)
    " Component Cooling Water (CCW)
    " Intake Cooling Water (ICW)
    " Residual Heat Removal (RHR)
     " Reactor Coolant System (RCS)
     " Main Steam (MS)
     " Chemical and Volume Control System (CVCS)
     " High Head Safety Injection (HHSI)
     " Instrument Air (IA)
Consensus was achieved in the expert panel process by discussing individual scenarios, reaching a conclusion, and asking for any dissenting opinions.
Follow-on Expert Panel 2010 The follow-on expert panel meeting in October 2010 utilized the same information, but with an updated PWROG Generic MSO list from NEI 00-01 Revision 2 and current lessons learned from the MSO process in FAQ 07-0038 and NFPA 805 pilot plants.
A brief overview training was conducted since all of the participants were familiar with the process and had previously participated in the earlier MSO expert panel meetings.
Page F-4 Revision 0 Revision  0                                                                        Page F-4

Florida Power & Light                             Attachment F - Fire-Induced MSOs Resolution The follow-on meeting in 2010 consisted of a review of outstanding action items and a review of items that had been added or changed from the PWROG generic MSO list from NEI 00-01, Revision 2.
Consensus was achieved in the expert panel process by discussing individual scenarios, reaching a conclusion, and asking for any dissenting opinions.
Step 3 - Update the Fire PRA model and NSCA to include the MSOs of concern.
This includes the:
    " Identification of equipment (NUREG/CR-6850 Task 2)
    " Identification of cables that, if damaged by fire, could result in the spurious operation (NUREG/CR-6850 Task 3, Task 9)
    " Identify routing of the cables identified above, including associating that routing with fire areas, fire zones and/or Fire PRA physical analysis units, as applicable.
Include the equipment/cables of concern in the Nuclear Safety Capability Assessment (NSCA). Including the equipment and cable information in the NSCA does not necessarily imply that the interaction is possible since separation/protection may exist throughout the plant fire areas such that the interaction is not possible).
Note: Instances may exist where conditions associated with MSOs do not require update of the Fire PRA and NSCA analysis. For example, Fire PRA analysis in NUREG/CR-6850 Task 2, Component Selection, may determine that the particular interaction may not lead to core damage, or pre-existing equipment and cable routing information may determine that the particular MSO interaction is not physically possible.
In other instances, the update of the PRA may not be warranted if the contribution is negligible. The rationale for exclusion of identified MSOs from the Fire PRA and NSCA should be documented and the configuration control mechanisms should be reviewed to provide reasonable confidence that the exclusion basis will remain valid.
Results of Step 3:
The NSCA and Fire PRA were updated to reflect the treatment of applicable MSO scenarios. This included the identification of equipment, identification of cables, and the routing of cables by plant locations. The PTN Results are documented in:
    " NSCA - PTN Post-Fire Safe Shutdown Analysis (report entitled, NFPA 805 Fire Area Transition Review)
    " Fire PRA - PTN Fire PRA Documentation (report entitled, Turkey Point Units 3&4 Fire Probabilistic Risk Assessment Component and Cable Selection Report)
Step 4 - Evaluate for NFPA 805 Compliance The MSO combinations included in the NSCA should be evaluated with respect to compliance with the deterministic requirements of NFPA 805, as discussed in Section 4.2.3 of NFPA 805. For those situations in which the MSO combination does not meet the deterministic requirements of NFPA 805 (VFDRs), the issue with the components and associated cables should be mitigated by other means (e.g., performance-based approach per Section 4.2.4 of NFPA 805, plant modification, etc.).
Page F-5 Revision 00 Revision                                                                              Page F-5

Florida Power & Light                           Attachment F - Fire-Induced IVISOs Resolution The performance-based approach may include the use of feasible and reliable recovery actions. The use of recovery actions to demonstrate the availability of a success path for the nuclear safety performance criteria requires that the additional risk presented by the use of these recovery actions be evaluated (NFPA 805 Section 4.2.4).
Results of Step 4:
The MSO combination components of concern were also evaluated as part of the PTN NSCA. For cases where the pre-transition MSO combination components did not meet the deterministic compliance, the MSO combination components were added to the scope of the fire risk evaluations. The process and results for Fire Risk Evaluations are summarized in Section 4.5 of the Transition Report.
Step 5 - Document Results The results of the process should be documented. The results should provide a detailed description of the MSO identification, analysis, disposition, and evaluation results (e.g., references used to identify MSOs; the composition of the expert panel, the expert panel process, and the results of the expert panel process; disposition and evaluation results for each MSO, etc.). High level methodology utilized as part of the transition process should be included in the 10 CFR 50.48(c) License Amendment Request/Transition Report.
Results of Step 5:
The PTN Results are documented in:
* MSO Expert Panel Report (report entitled, Expert Panel for Addressing Multiple Spurious Operations)
* NSCA - PTN Post-Fire Safe Shutdown Analysis (report entitled, NFPA 805 Fire Area Transition Review)
* Fire PRA - PTN Fire PRA Documentation (report entitled, Turkey Point Units 3&4 Fire Probabilistic Risk Assessment Component and Cable Selection Report)
* PTN Fire Risk Evaluations - See NEI 04-02 Table B-3 (Attachment C-1)
Page F-6 Revision 00                                                                         Page F-6

Florida Power & Light                         Attachment H - NEI 04-02 FAQs Summary Table H. NFPA 805 Frequently Asked Question Summary Table 2 Pages Attached Note: The NFPA 805 FAQ process will continue through the transition of non-pilot NFPA 805 plants. Final closure of the FAQs will occur when RG 1.205, which endorses the new revision of NEI 04-02, is approved by the NRC.
Page H-I Revision 00                                                                     Page H-1

Florida Power & Light                                   Attachment H - NEI 04-02 FAQs Summary Table This table includes the approved FAQs that have not been incorporated into the current endorsed revision of NEI 04-02 and utilized in this submittal:
Table H NEI 04-02 FAQs Utilized in LAR Submittal No.      Rev.                      Title                     FAQ Ref.            Closure Memo 06-0008      9     NFPA 805 Fire Protection Engineering       ML090560170         ML073380976 Evaluations 06-0022      3     Acceptable Electrical Cable                ML090830220         ML091240278 Construction Tests 07-0030      5     Establishing Recovery Actions              ML103090602         ML110070485 07-0032      2     Clarification of 10 CFR 50.48(c),          ML081300697         ML081400292 10 CFR 50.48(a) and GDC 3 clarification 07-0035            Bus Duct Counting Guidance for High        ML091610189         ML091620572 Energy Arcing Faults Lessons learned on Multiple Spurious                           ML110140242 07-0038                                                       ML103090608 Operations 07-0039            Lessons Learned - NEI B-2 Table            ML091420138         ML091320068 07-0040            Non-Power Operations Clarification         ML082070249         ML082200528 08-0042            Fire Propagation from Electrical           ML080230438         ML092110537 Cabinets                                   ML091460350 08-0043            Electrical Cabinet Fire Location           ML083540152          ML092120448 ML091470266 08-0044            Large Oil Fires                            ML081200099          ML092110516 ML091540179 08-0046            Incipient Fire Detection Systems           ML081200120          ML093220426 ML093220197 08-0047            Spurious Operation Probability             ML082770662          ML082950750 08-0048            Fire Ignition Frequency                    ML081200291          ML092190457 ML092180383 08-0049            Cable Fires                                ML081200309          ML092100274 ML091470242 08-0050            Non Suppression Probability                ML081200318          ML092190555 ML092510044 08-0051            Hot Short Duration                         ML083400188          ML100900052 ML100820346 08-0052           Transient Fire Growth Rate and             ML081500500          ML092120501 Control Room Non-Suppression               ML091590505 Page H-2 Revision 0 Revision                                                                                     Page H-2

Florida Power & Light                                    Aftachment H - NEI 04-02 FAQs Summary Table Table H NEI 04-02 FAQs Utilized in LAR Submittal No.        Rev.                      Title                       FAQ Ref.               Closure Memo 07-0054*        1      Demonstrating Compliance with              ML103510379           ML110140183 Chapter 4 of NFPA 805 09-0056        2      Radioactive Release Transition             ML102810600           ML102920405 09-0057        3      New Shutdown Strategy                      ML100330863           ML100960568 10-0059        5      NFPA 805 Monitoring                        ML120410589           ML120750108 12-0062        0      UFSAR Content                              ML120790015
* Note: The FAQ submittal number was 08-0054 but the NRC closure memo for the FAQ was listed as 07-0054. 07-0054 was used to be consistent with the Closure Memo.
Page H-3 Revision 0                                                                                          Page H-3

Florida Power & Light                        Attachment I- Definition of Power Block
: 1. Definition of Power Block I Page Attached Page I-I Revision 00 Revision                                                                     Page 1-1

Florida Power & Liaht                                               Attachment I - Definition of Power Block For the purposes of establishing the structures included in the Fire Protection program in accordance with 10 CFR 50.48(c) and NFPA 805, plant structures listed in the following table are considered to be part of the power block.
Table I Power Block Definition Power Block Structures                                      Fire Area(s)/ Zone(s)
Unit 3 Containment              Q/60 Unit 4 Containment              P/59 Auxiliary Building             A/4-10,17-18; B/11-13; C/14-16; D/30, E/40; F/31-39,48-51,58; G/25-25A; H/19; 1/20; J/26; K/27; L/28; N/45; 0/41,55; AAA/1-3,21-24,46,111-112; BBB/52-53; FFF/107; OD/47,54,118,123 Unit 3 Fuel Handling Building        AAA/42,56-57 Unit 4 Fuel Handling Building        AAA/29,43-44 Control Building              R/61; S/62; T/63; HH/98, 132; 11/101; JJ/102; KK/103; LL/104; MM/97,106,106R; NN/108A; 00/108B; PP/109; QQ/110; AAA/65; CCC/64 Turbine Building              U/67; V/68; W/70; X/71; DD/93; EE/94; FF/95; GG/96; DDD/66; EEE/69; OD/76-78,80-83,85-87,91-92,105,117 Outdoor Area East of Turbine Building    CC/79A; OD/79,84,88-89,113-116 Intake Structure              OD/119-122 Unit 3 EDG Building             Y/72; Z/73; AA/74; BB/75; OD/90,131,143 Unit 4 EDG Building             RR/138; SS/133; TT/134; UU/139; VV/140; WW/141-142; YY/135; XX/1 36-137 Unit 3 C-Bus Switchgear Building       OD/999 (151)
Unit 4 C-Bus Switchgear Building       OD/999 (152)
Unit 3 Condensate Polishing Building     OD/999 (151)
Unit 4 Condensate Polishing Building     OD/999 (152)
Unit 3 Area NE of Equipment Hatch       OD/124 Unit 4 Area SE of Equipment Hatch       OD/125 Radwaste Building              AAA/126A-E Hydrogen Storage Building          OD/999 (F12)
Note: For Fire Area OD, Fire Zone 999 is defined as Miscellaneous Areas; some of the structures within Fire Zone 999 have been assigned unique fire zones. These unique zones are identified in parentheses in the above table.
Page 1-2 Revision 00 Revision                                                                                                   Page 1-2

Florida Power & Light                      Affachment J - Fire Modelinq V&V Flrd oe     ih                        AfahetJ-FieMdln               &
J. Fire Modeling V&V 23 Pages Attached Page J-1 Revision 00                                                         Page J-1

Florida Power & Light                                        Affachment J - Fire Modeling V&V This attachment documents the Verification and Validation (V&V) basis for the Fire Probabilistic Risk Assessment (FPRA) fire modeling applications at Turkey Point Nuclear (PTN). Plant specific fire modeling used to support the PTN FPRA consists of the following:
     " The calculation of the main control room operator abandonment times [Report 0020-0010-000, Rev. 0]; and
     " The use of generic fire modeling treatments and its associated supplements as applicable to develop Zones of Influence (ZOI) [Hughes Associates, 2008; Hughes Associates, 2011].
Main Control Room Abandonment Report The goal of the main control room abandonment report titled "Evaluation of Main Control Room Abandonment times at the Turkey Point Nuclear Plant" [0020-0010-000, Rev. 0]
is to compute the time operators would abandon the main control room for both Units 3 and 4 using the NUREG/CR-6850 [2005] abandonment criteria for control room fire scenarios. The abandonment times are assessed for various electronic equipment fires and for ordinary combustible fires as defined by the discretized heat release rate conditional probability distributions presented in NUREG/CR-6850 [2005]. The abandonment time in the main control room is estimated by calculating the time to reach threshold values for temperature and visibility as identified by NUREG/CR-6850 [2005].
An additional abandonment criterion based on NUREG-0700 [2002] and the Society of Fire Protection Engineers (SFPE) Handbook of Fire Protection Engineering, Section 2-6 [2002] that is more conservative than the NUREG/CR-6850 [2005] criteria. This is included to account for potential abandonment due to immersion of the operators in an elevated temperature environment.
The focus of the MCR abandonment evaluation is on the first twenty-five minutes after ignition because the non-suppression probability (NSP) decreases to 0.001 at twenty minutes [NUREG/CR-6850, 2005]. The abandonment calculations are performed using the zone fire model Consolidated Fire and Smoke Transport (CFAST), Version 6.0.10.61027 [National Institute of Standards and Technology (NIST) Special Publication (SP) 1026, 2009 and NIST SP 1041, 2008].
The main control room area geometry and fire parameters for the simulations fall within the model limits listed in NIST SP 1026 [2009], and NIST SP 1041 [2008]. Specifically, the vent area to enclosure volume ratio is less than two and the aspect ratio of the enclosures is less than five (for the true geometry). The physical input dimensions are adjusted to account for obstructions and boundary heat losses and the resulting model geometry has a length to width aspect ratio greater than five. However, the input geometry conserves the boundary area, room volume, and enclosure height and a corridor flow model is intentionally avoided because the true geometry has an aspect ratio that is within the model limitations.
The non-dimensional parameters that affect the model results as documented in NUREG-1824, Volumes 1 and 5 and NUREG-1934 [2011] include the model geometry, the global equivalence ratio, the fire Froude Number, and the flame length ratio. Non-dimensional parameters that relate to target exposure conditions (heat flux) and Revision 0                                                                            Page J-2

Florida Power & Light                                          Aftachment J - Fire Modeling V&V sprinkler actuation (ceiling jet) are not applicable to this calculation because these output parameters are not used.
The non-dimensional geometry parameters (length to height and width to height, which range from 3.3 -4.3) fall within the NUREG-1824 validation range (0.6 - 5.7). CFAST does not use a fire diameter; therefore, it is possible to specify a fire that falls within the range of fire Froude numbers considered in the NUREG-1824 validation documentation.
The source fires considered are consistent with those described in NUREG/CR-6850
[2005] and thus are the subject of the NUREG-1824 validation effort. The global equivalence ratio for normal supply may be assessed using the ratio of the maximum supported fire size to the fire size postulated. Based on the supply flow of 4.8 m3 (10,160 cfm), the maximum fire size that could be supported is about 17 MW (16,100 Btu/s). The maximum fire size postulated is about 5 MW (4,700 Btu/s); thus the global equivalence ratio is less than 0.3 and falls within the NUREG-1824 validation range. In the case of an inoperative ventilation system, the abandonment times correspond to a point on the timeline when the global equivalence ratio is significantly less than unity and thus within the NUREG-1824 range. The initial reservoir of oxygen within the enclosure supports the postulated fires sufficiently long for abandonment to occur prior to the onset of ventilation limited burning. Finally, the flame length ratio is normally met, but in the case of the largest fire sizes postulated, the flame height may reach or exceed the ceiling height. Because sprinkler actuation and thermal radiation to targets are not computed with the CFAST model, this parameter is not an applicable metric. Rather, the plume entrainment below the hot gas layer controls the layer descent time and the concentration of soot products in the layer. This aspect of the model is not affected by the flame height to ceiling height ratio. Consequently, the application of CFAST at the PTN control room falls entirely within the NUREG-1824, Volume 1 [2007] V&V space.
Additional verification V&V studies are contained in NIST SP 1086 [2008] and Tatem et al. [2004]. These studies have a broader parameter V&V space than NUREG-1824, Volume 1 [2005] and NIST SP 1086 [2008] is based in part on the methods of American Society for Testing and Materials (ASTM) El 355 [2004]. Tatem et al. [2008] provides a Navy specific V&V study, which includes an assessment of CFAST, Version 3.1.7 predictions in multiple enclosures and multiple elevation configurations. These additional V&V studies extend the range of the V&V space to include configurations and conditions applicable to the MCR abandonment sensitivity analysis (Appendix B).
The main control room abandonment report also provides benchmark and validation simulations for CFAST as applicable to the main control room area of PTN. In particular, the control room tests documented in NUREG/CR-4527, Volume 2 [1988],
are used to provide additional validation basis for control room application of CFAST.
Table J-1 provides a summary of the validation and verification basis for CFAST, Version 6.0.10.61027 as applied in the main control room abandonment report.
Generic Fire Modeling Treatments Report The generic fire modeling treatments, Revision 0 is used to establish zones of influence for specific classes of ignition sources and primarily serves as a screening calculation in the FPRA under NUREG/CR-6850 [2005], Sections 8 and 11. The generic fire modeling treatments document has two fundamental uses within the FPRA:
Revision 0                                                                               Page J-3

Florida Power & Light                                       Aftachment J - Fire Modeling V&V
    "   Determine the ZOI inside which a particular ignition source is postulated to damage targets; and
    " Determine the potential of the ignition sources to generate a hot gas layer within an enclosure that can either lead to full room burnout or invalidate the generic treatment ZOls for a particular class of combustible materials.
The ZOI is determined using a collection of empirical-and algebraic models and correlations. The potential for a hot gas layer having a specified temperature to form within an enclosure is determined using the zone model CFAST, Version 6.0.10.61027
[NIST SP 1026, 2009 and NIST SP 1041, 2008].
Verification The calculation development and review process in place at the time the Generic Fire Modeling Treatments was prepared included contributions from a calculation preparer, a calculation review, and a calculation approver. The responsibilities for each are as follows:
    " The calculation prepare develops and prepares the calculation using appropriate methods.
     " The calculation reviewer provides a detailed review of the report and supporting calculations, including spreadsheets and fire model input files. The reviewer provides comments to the preparer for resolution.
     " Calculation approver provides a reasonableness review of the report and approves the document for release.
The calculation preparation occurred over a two year period ending in 2007. The review stage was conducted in 2007 at the completion of the preparation stage. The calculation was approved January 23, 2008. The approved document, the signature page, and an affidavit were transmitted to the Document Control Desk at the Nuclear Regulatory Commission in Washington, D. C. on January 23, 2008.
In the case of the empirical equations/correlations that form part of the basis of the "Generic Fire Modeling Treatments," a considerable amount of verification was performed during the preparation stage by the preparer. The empirical equations/correlations were solved using Excel spreadsheets using either direct cell solutions (algebraic manipulation) or Visual Basic macros. All direct cell solutions were validated by the preparer through the use of alternate calculation. For simple equations, this entailed matching spreadsheet solution to the solution obtained using a hand calculator. For more complex solutions, the alternate calculation verification entailed either subdividing the problem into many sub-components and matching the solution using a hand calculator or matching the solution to a verified solution (i.e., the NUREG-1805 [2004] Solid Flame Heat Flux models). The verification of the Visual Basic macros also depended on the type of macro. In situations where the macro is used to perform multiple direct computations, the macro results were verified against the verified spreadsheet solutions that were verified through alternate calculation. In cases where the macro is used to find a root, the root was verified to be a zero by direct substitution into an alternate form of the solved equation.
Revision 0                                                                           Page J-4

Florida Power & Liqht                                        Aftachment J - Fire Modelina V&V The empirical equations/correlations were further verified by the reviewer using a Design Review method as indicated in the signature sheet. An independent reviewer was provided access to the draft report and all supporting calculation materials in late 2007. The reviewer conducted a detailed review of the implementation of the equations within the spreadsheets and the reporting of the equation result in the draft report.
Comments and insights were provided to the preparer over the review period and were addressed to the satisfaction of the reviewer. Upon the completion of the review, a revised draft was prepared for review by the approver. The approver provided a higher level reasonableness check of the methods, approach, and the results. Comments and insights that were provided by the approver were addressed to the satisfaction of the reviewer and Rev. 0 of the report was prepared and approved on January 23, 2008.
Validation The empirical equations and correlations are drawn from a variety of sources that are documented in various chapters of the Society of Fire Protection Engineers Handbook of Fire ProtectionEngineering, peer reviewed journals (e.g., the Fire Safety Journal), or engineering textbooks. The empirical models primarily fall into three groups:
     " Flame height;
* Plume temperatures; and
    " Heat fluxes (at a target location).
Table J-2 of this attachment identifies the empirical models that are used either directly or indirectly in the "Generic Fire Modeling Treatments" report. The table also identifies the original correlation source documentation and the correlation range in terms of non-dimensional parameters. The table also provides where applicable supplemental validation work that may have been performed on the correlations and provides limits applied in the "Generic Fire Modeling Treatments" report as applicable.
Except for the cable tray Zone of Influence (ZOI) calculation, the flame height calculation is used only as a means of placing a limit on the applicability of the ZOI tables which are based on the plume temperature and thermal radiation heat flux. The flame height calculation for axisymmetric source fires is robust and has considerable pedigree. The original documentation and basis of the flame height correlation is Heskestad [1981] as noted in Table J-2 of this attachment. Although there are earlier forms of the flame height equation, Heskestad provides a link between the flame height and plume centerline temperature calculation and identifies the range over which the plume equations are applicable. Because the flame height and plume centerline temperature equations are linked, the plume centerline range cited by Heskestad applies to the flame height calculation as well. The plume centerline temperature equations, and thus the flame height correlation, is applicable over the following range as noted in Table J-2 [Heskestad, 1981; Heskestad, 1984]:
5 iOg~o 10 [1(A.i/T)r  ).]     S                        (J-1) where cp is the heat capacity of ambient air (kJ/kg-K [Btu/Ib-&deg;R]), To is the ambient temperature (K ['R ]), g is the acceleration of gravity (m/s 2 [ft/s 2 ]), pc is the ambient air Revision 0                                                                                Page J-5

Florida Power & Light                                          Attachment J - Fire Modeling V&V density (kg/M 3 [Ib/ft]), Q is the fire heat release rate (kW [Btu/s), r is the stoichiometric fuel to air mass ratio, D is the fire diameter (m [ft]), and AH- is the heat of combustion of the fuel (kJ/kg [Btu/Ib]). Application of Equation (1) depends on the fuel as well as a non-dimensional form of the fire heat release rate (fire Froude Number). In practice, the heat of combustion to air fuel ratio for most fuels will fall between 2,900 - 3,200 kJ/kg (1,250 - 1,380 Btu/Ib), and for typical ambient conditions the D5   L ratio for which the plume equations have validation basis is between 7 - 700 kW 215/m (4 - 9 Btu 2/5/ft)
[Heskestad, 1984]. For fire sizes on the order of 25 kW (24 Btu/s) or greater, this 2
means that the plume centerline equation is valid for heat release rates of 100 kW/m (8.81 Btu/s-ft 2) to well over 3,000 kW/m 2 (264 Btu/s-ft 2). For weaker fires (heat release rates less than 100 kW/m 2 (8.81 Btu/s-ft2), the tendency of the model is clearly to over-predict the temperature and flame height; thus for applications outside the range but below the lower limit the result will be conservative. The concern is therefore entirely on the upper range of the empirical model. The tables in the "Generic Fire Modeling Treatments" are specifically developed with transient, lubricant spill fires, and electrical panel fires with a heat release rate per unit area within the validation range. When the heat release rate per unit area falls outside the applicable range, the table entry is not provided and it is noted that the source heat release rate per unit area is greater than the applicable range for the correlations. This applies to the flame height and the plume temperature for axisymmetric source fires.
The flame height and plume centerline temperature for line type fires (fires having a large aspect ratio) are applied only to cable tray fires. The correlation used has pedigree and has existed in its general form since at least Yokoi [1960]. Most recently, Yuan et al. [1996] provided a basis for the empirical constant using experimental data with source fires having a width of 0.015 m -0.05 m (0.05- 0.15 ft) and a length of 0.2
- 0.5 m (0.7 - 1.5 ft) [Yuan et al., 1996]. When normalized, the applicable height to heat release rate per unit length range (4) for the correlations based on the experiments Q
of Yuan et al. [1996] is between 0.002 and 0.6. This range includes the flame height as well as the elevation at which the temperature is between 204 - 3290C (400 - 625&deg;F),
the temperature at which cable targets are considered to be damaged under steady state exposure conditions. Yuan et al. [1996] also provide a tabular comparison of the empirical constant against seven preceding line fire test series, which include a broader range of physical fire sizes and dimensions. The Yuan et al. [1996] constant is greater than the other seven and thus the temperatures and flame heights are more conservatively predicted using the Yuan et al. [1996] data. The application of the Yuan et al. [1996] correlation in the "Generic Fire Modeling Treatments" falls within the normalized applicability range reported by Yuan et al. [1996].
Four flame heat flux models are used in the "Generic Fire Modeling Treatments" as described in Table J-2 of this attachment: the Point Source Model, the (simple) Method of Shokri and Beyler, the Method of Mudan and Croce, and the Shokri and Beyler Method. The former two are simple algebraic models using the heat release rate, separation distance, and the fire diameter. The latter two are considered detailed radiant models that account for the emissivity of the fire and the shape of the flame.
Due to limitations in the target placement, the (Simple) Method of Shokri and Beyler are Revision 0                                                                              Page J-6

Florida Power & Light                                            Affachment J - Fire Modeling V&V shown to be inapplicable for calculating the ZOI dimensions. Similarly, for the fuels considered, it is shown that the Method of Mudan and Croce produce a net heat flux that exceeds the fire size. The ZOls are therefore determined using the Point Source Model and the Method of Shokri and Beyler. The method that produces the largest ZOI dimension is used for each fuel and fire size bin.
The Point Source Model and the Method of Shokri and Beyler have been shown in the NUREG-1824 verification and validation study to provide reasonably accurate predictions when the target separation to fire diameter ( -R) ratio is between 2.2 and 5.7
[NURE-1824, Volume 1, 2007]. Furthermore, the fire size ranges considered in the "Generic Fire Modeling Treatments" report are between about 25 - 12,000 kW (24 -
11,400 Btu/s) and the heat release rates per unit area range between about 100 - 3,000 kW/m 2 (8.1 - 264 Btu/s-ft2) for all fuels and fire size bins. Using this information, the following table may be assembled for the applicable target heat flux range, based on the NUREG-1824, Volume 1 [2007] verification and validation range:
Heat Release Rate    Fire Diameter    Point Source Model   Shokri and Beyler Per Unit Area                         Heat Flux Range     Heat Flux Range (kW [Btus])              [Btu/sftA )      (m [ft])      (kW/m2 [Btu/s-ft2]) (kW/m 2 [Btu/s-ft2])
0.07-0.45            0.36- 3.8 25(24)             100(8.8)           0.56(1.9)         (0.006- 0.04)        (0.03-0.4) 2-13.6             2.84-10 25(24)           3,000 (264)          0.1 (0.3)           (0.2-1.2)           (0.3-0.9) 0.07 -0.45            0.55-12,000 (11,400)         100 (8.8)          12.4(41)          (0.006-0.04)         (0.05 - 0.4) 2-13.6             0.45 -4.6 12,000 (11,400)       3,000 (264)           2.3 (7.4)          (0.2 - 1.2)        (0.04-0.4)
The threshold heat fluxes that define the steady state ZOI dimensions range from 5.7 -
11.4 kW/m 2 (0.5 - 1 Btu/s-ft2). Transient ZOI dimensions, addressed in the "Supplemental Generic Fire Modeling Treatments: Transient Ignition Source Strength" may approach 16 - 18 kW (1.4 - 1.6 Btu/s-ft2 ). Clearly, the steady state ZOI dimensions based on critical heat fluxes of 5.7 - 11.4 kW/m 2 (0.5 - 1 Btu/s-ft2 ) overlay with the range of valid predicted heat fluxes identified in NUREG-1824, Volume 1
[2007]. Fuels that identify the most conservative value over a range of heat release rates per unit area (transient and electrical panels) will thus include at least one point within the validation range (i.e., 5.7 kW/m 2 [0.5 Btu/s-ft2 ]). Since the algorithm searches for the most adverse value, the result will be at least as conservative as the value obtained within the model validation and verification range.
There are combinations of fuels and source strength ranges that do not produce heat fluxes that fall within the validation range. This is especially true for the higher target heat flux values (11.4 kW/m 2 [1 Btu/s-ft2] and higher) combined with the lower transient fuel 2
package heat release per unit area range (200 - 1,000 kW/m2 [17.6-88.1 Btu/s-ft ]). This is addressed through an extended Verification and Validation range of the heat flux models provide by the SFPE [SFPE, 1999]. As noted in Table J-2 of this attachment, the SFPE assessed the predictive capabilities of the Point Source Model Revision 0                                                                                 Page J-7

Florida Power & LiQht                                         Attachment J - Fire Modeling V&V and the Method of Shokri and Beyler against available pool fire data. The pool diameters ranged from 1 - 80 m (3.3 - 262 ft). The conclusion was that the Point Source Model was conservative, but not necessarily bounding, when the predicted heat flux is less than 5 kW/m 2 (0.44 Btu/s-ft2 ) and the empirical constant (radiant fraction) is 0.21. The method is bounding when a safety factor of two is applied to the predicted heat flux. The application in the "Generic Fire Modeling Treatments" uses an empirical constant (radiant fraction) of 0.35, indicating the application is essentially bounding.
Similarly, it was concluded that that Method of Shokri and Beyler is conservative when the predicted heat flux is greater than 5 kW/m 2 (0.44 Btu/s) and the method is bounding when a safety factor of two is applied to the predicted heat flux. The implementation in the "Generic Fire Modeling Treatments" is conservative, though not bounding. Although the SFPE considered fire diameters greater than about 1 m (3.3 ft), smaller diameter pool fires are not optically thick and have a lower emissive power [SFPE Handbook of Fire Protection Engineering, Section 3-1, 2008]. Thus, the use of the methods for smaller fires is conservative though outside the SFPE validation range.
The use of the heat flux models largely falls within the NUREG-1824, Volume 1 [2007]
verification and validation parameter space range; however there are cases where this is not so. However, for larger diameter fires, the SFPE provides comprehensive validation against full scale test data of the methods applied. The application in the "Generic Fire Modeling Treatments" report and the applicable supplements necessarily fall within the validation range or are more conservative because the solution algorithm identifies the most adverse solution among the methods. Smaller fires may fall outside the validation range of both studies, but such fires have a lower emissive power and are conservatively treated using the methods designed for high emissive power source fires.
A number of other empirical models that appear in the generic fire modeling treatments are applied within the stated range of the models or the data for which the models were developed. For example, the cable heat release rate per unit area model is based on cables that have a small scale heat release rate that ranges between 100 - 1,000 kW/m 2 (8.8 - 88.1 Btu/s-ft2 ). The solution tables are provided for this range. The unconfined spill fire model (heat release rate reduction factor) is based on observations of pool fires having a diameter between 1 - 10 m (3.3 - 33 ft). The diameter range for which ZOI data is provided is 0.7 - 5 m (2.2 - 17 ft). The lower range value is less of a concern due the reduction in the optical thickness of the fire when the diameter falls below 1 m (3.3 ft). The upper range is maintained in the ZOI solutions. The offset distance for flame extensions outside a burning panel have an upper observational limit of about 1,000 kW (950 Btu/s), though it is applied in a normalized form (extension to panel height ratio). The ratio is applied as determined from the test data.
The CFAST applications in the "Generic Fire Modeling Treatments" report consist of simple geometries with a single natural vent path connected to an ambient boundary condition. The simulations are used to determine the time after the start of the fire that the hot gas layer temperature reaches a predetermined critical temperature. No consideration for the hot gas layer depth is made; if the hot gas layer temperature reaches the critical temperature at any time, then this time is the sole output parameter used in the "Generic Fire Modeling Treatments" report. The enclosure geometry is Revision 0                                                                            Page J-8

Florida Power & Light                                            Attachment J - Fire Modeling V&V specified as a function of the volume in such a way as to minimize the heat losses to the boundary. Three vent configurations are evaluated for each volume-room geometry-vent fraction; the most adverse result among the three vent configurations is used.
The room geometry and fire parameters for the "Generic Fire Modeling Treatments" simulations fall within the model limits listed in NIST SP 1026 [2009], and NIST SP 1041
[2008]. Specifically, the vent area to enclosure volume ratio is less than two and the aspect ratios of the enclosures are less than five.
The non-dimensional parameters that affect the model results as documented in NUREG-1824, Volumes 1 and 5 and NUREG-1934 [2011] include the model geometry, the global equivalence ratio, the fire Froude Number, and the flame length ratio. The non-dimensional parameters that relate to target exposure conditions (heat flux) and sprinkler actuation (ceiling jet) are not applicable to this calculation because these output parameters are not used. The non-dimensional geometry parameters (length to height and width to height, which range from 3.3 - 4.3) fall within the NUREG-1824 validation range (0.6 - 5.7). As previously noted, CFAST does not use a fire diameter; therefore, it is possible to specify a fire that falls within the range of fire Froude numbers considered in the NUREG-1824 validation documentation. The source fires considered are consistent with those described in NUREG/CR-6850 [2005] and thus those that are the subject of the NUREG-1824 validation effort. The global equivalence ratio does exceed the ratio validated in NUREG-1824, Volume 1, in some cases by a significant margin. Large fires in very small volumes with low ventilation could effectively result in equivalence ratios that even exceed the maximum values observed in fully developed fires (3 - 5) [SFPE Handbook of Fire Protection Engineering, Section 2-5, 2008; SFPE Handbook of Fire Protection Engineering, Section 3-4, 2008]. However, the limiting oxygen index used in the model is zero, which forces the combustion process to use all available oxygen within the enclosure and the heat release rate to decrease to a value set by the natural ventilation oxygen inflow. The maximum temperature over the course of the fire occurs at some time prior to the oxygen being consumed in the enclosure, thus the global equivalence ratio for the data reported is based on a condition where it is less than unity and within the validation basis of NUREG-1824, Volume 1 [2007].
Further, for a given volume and fire size, an optimum ventilation condition will occur over the vent range considered. Because of potential variations in a ventilation condition, the FPRA uses the most adverse time over the reported range and effectively performs an optimization on this parameter.
Finally, the flame length ratio is not always met, especially for large fires postulated in small enclosures. Because sprinkler actuation and thermal radiation to targets are not computed with the CFAST model, this parameter is not an applicable metric. Rather, the plume entrainment below the hot gas layer controls the layer decent time and the concentration of soot products in the layer. This aspect of the model is not affected by the flame height to ceiling height ratio. Consequently, the application of CFAST at the PTN control room falls entirely within the NUREG-1824, Volume 1 [20071 V&V parameter space.
Additional verification V&V studies are contained in NIST SP 1086 [2008] and Tatem et al. [2004]. These studies have a broader parameter V&V space than NUREG-1824, Volume 1 [2005] and NIST SP 1086 [2008] is based in part on the methods of American Revision 0                                                                                Page J-9

Florida Power & Light                                         Aftachment J - Fire Modeling V&V Society for Testing and Materials (ASTM) El 355 [2004]. Tatem et al. [2008] provides a Navy specific V&V study, which includes an assessment of CFAST, Version 3.1.7 predictions in multiple enclosures and multiple elevation configurations. These additional V&V studies extend the range of the V&V parameter space to include configurations and conditions applicable to the MCR abandonment sensitivity analysis (Appendix B of the "Generic Fire Modeling Treatments" report).
Appendix B of the "Generic Fire Modeling Treatments" report provides an in depth analysis of the parameters used as input and Table B-2 indicates the basis for the input parameter selection. The parameters are either selected as absolutely bounding over the credible range or establish an application limit (e.g., elevated temperature environment and boundary thermal properties).
A summary of the validation basis for both the CFAST and the empirical models is provided in Tables J-1 and J-2 of this attachment. Based on the information in the tables and the preceding discussion, it is shown that that the empirical fire model applications in the "Generic Fire Modeling Treatments" either fall within the original correlation bounds or they are outside the bounds but used in a way that is demonstrably conservative. Likewise, CFAST is used within the model limitations described in the User's Guide [NIST SP 1041, 2008] and the Technical Reference Guide [NIST SP 1026, 2009]. The results as reported in the "Generic Fire Modeling Treatments" document are based on conditions that meet the NUREG-1824, Volumes 1 and 5 V&V space, although there are input specifications that fall outside this range.
The use of the "Generic Fire Modeling Treatments" in the FPRA performs an optimization over the ventilation fraction and necessarily is based on a condition that falls within the NUREG-1 824, Volumes 1 and 5 V&V space for the global equivalence ratio.
Generic Fire Modeling Treatments Supplements There are four supplements to the "Generic Fire Modeling Treatments", three of which are used by the PTN FPRA:
     " Supplement 1: "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Revision B;
     " Supplement 2: "Supplemental Generic Fire Modeling Treatments: Hot Gas Layer Tables," Revision G; and
     " Supplement 3: "Supplemental Generic Fire Modeling Treatments: Transient Ignition Source Strength," Revision E.
Supplement 4, "Supplemental Generic Fire Modeling Treatments: Transient Target Response to Transient Ignition Source Fire Exposures," Revision A is not used by the PTN FPRA.
Supplement I Supplement 1, "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Revision B considers the maximum potential fire size that can develop in electrical panels that are vented. The supplement consists of a theoretical model development (energy and mass balance on a control volume) and a parametric sensitivity analysis. The sensitivity analysis consists of two parts: the first part provides Revision 0                                                                            Page J-10

Florida Power & Light                                         Affachment J - Fire Modeling V&V the maximum heat release rate that can be supported by a vented metal box and the second part uses this heat release rate information as input into the same calculation process used by the "Generic Fire Modeling Treatments" report to generate the ZOI dimensions. The numerical solution for the heat release rate calculation was independently reviewed in the same manner as was the original "Generic Fire Modeling Treatments" calculations. The ZOI dimension calculation is the same procedure used in the original "Generic Fire Modeling Treatments" report and was verified and approved as previously described.
An alternate method verification is also provided in Section 6 of "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Revision B. The alternate method for which the model is compared to was developed by Electric Power and Research Institute and documented in a draft EPRI report [EPRI, 2011]. The alternate calculation comparison only considers the ventilation limited aspect of the EPRI model. No consideration of the heat release rate conditional probability distribution is provided in "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Revision B.
The model is validated against available test data for either electrical panel fire tests or tests in metal enclosures having dimensions comparable to electrical panels. These tests are documented in the following references:
* Mangs et al., 1994.
* Mangs et al., 1996.
* Mangs et al., 2003.
* Mangs et al., 2004.
* Melis et al., 2004.
* NUREG/CR-4527, Volume 1, 1987.
* NUREG/CR-4527, Volume 2,1988.
Approximately thirty fire tests are represented in the references cited. A common attribute among the tests was the desire to drive the internal conditions to a fully-developed (i.e., ventilation controlled) burning regime. Not all tests were successful, and consequently a fully developed model will over-predict the heat release rate. It is shown by Hunt [2009] that the fully developed model predicts or over-predicts the heat release rate from all fire tests except for five tests. Four of the tests involved a polymethylmethacrylate (PMMA) fuel that exhibited combustion external to the panel at the vents [Melis et al., 2004]. The recommended procedure for addressing external burning at the vents is to assume a global equivalence ratio within the panel of two, with half of the combustion occurring inside the panel and half outside the panel. This applies only to fuels that are shown to re-ignite outside the panel and would not typically be assumed for cable materials. The fifth test for which the heat release rate was not conservatively predicted involved a change in the status of the opening (i.e., the door latch mechanism failed and the door opened). The recommended practice to account for this situation is to use the FAQ 08-0042 [NUREG/CR Supplement 1, 2010] guidance for determining whether a panel is robustly secured.
Page J-11 Revision 00 Revision                                                                             Page J-1 1

Florida Power & Light                                        Affachment J - Fire Modelinq V&V The mass and energy balance model used to calculate the maximum fire size that can be supported within an electrical panel is conservative and is shown to predict the heat release rate among a large number of fire tests. The tests consist of a large range of ventilation fractions and panel sizes. In addition, the input parameters are conservatively biased as described in Section 5 of "Supplemental Generic Fire Modeling Treatments: Closed Electrical Panels," Revision B. As such, the model is considered to be validated over the range for which the ZOI dimensions are tabulated in the supplement.
Supplement 2 Supplement 2, "Supplemental Generic Fire Modeling Treatments: Hot Gas Layer Tables," Revision G provides hot gas layer tables for additional critical temperatures, and ignition source heat release rates, including some ignition source-secondary fuel package combinations. In addition, the ZOI dimensions for sensitive component targets is provided. The hot gas layer tables and ZOI dimensions are calculated using the same calculation procedures as was used for the original "Generic Fire Modeling Treatments" report but with different input parameters. These procedures were verified and approved as previously described. The validation basis for the ZOI dimensions is identical to that of the original "Generic Fire Modeling Treatments" report.
Hot gas layers are provided for a single secondary combustible configuration: two horizontal cable trays located 0.3 m (1 ft) above an electrical panel ignition source. The cable trays are simultaneously ignited five minutes after ignition at a single point and the fire is allowed to propagate laterally in each direction. The propagation rate is as recommended in NUREG/CR-6850 [2005] for thermoplastic and thermoset cables. This propagation rate has been shown to be broadly applicable to the cable class in NUREG/CR-7010 [2011]. The assumed heat release rate per unit area for the cables is constant and assumed to be 225 kW/m 2 (19.8 Btu/s-ft2), which is slightly less than the value recommended in NUREG/CR-7010 for thermoplastic cables (250 kW/m2 [22 Btu/s-ft2]).
Supplement 3 Supplement 3, "Supplemental Generic Fire Modeling Treatments: Transient Ignition Source Strength," Revision E provides an analysis of the heat release rate per unit area for transient fuel packages using the original transient fire test data referenced in NUREG/CR-6850. The heat release rate per unit area is directly related to the fire area and fire diameter when the heat release rate is specified as an input. Given the sensitivity of the plume and thermal radiation models to the fire diameter, a narrower heat release rate per unit area range eliminates over-conservatism for some applications.
The analysis is not a fire model per se, but it was independently reviewed. The use of the revised heat release rate per unit area ranges was input into the same process that was used to generate the ZOls in the original "Generic Fire Modeling Treatments" report. The validation and verification basis for the ZOI dimensions using the revised heat release rate per unit area as an input is thus identical to that of the original "Generic Fire Modeling Treatments" report.
Page J-12 Revision 00                                                                           Page J-12

0 Florida Power & Light                                                                                                 Attachment J - Fire Modeling V&V Table J-1 V & V Basis for Fire Models I Model Correlations Used in FPRA Calculation                Application                         V & V Basis                                          Discussion Main CR Abandonment. Calculation of operator                 NUREG-1824, Volume 5           The abandonment time in the MCR is determined by abandonment times in the                      NIST SP 1026             computing the time for the visibility and temperature to reach Main Control Room.                           NIST SP 1041              thresholds as specified in NUREG/CR-6850 [2005].
CFAST, version 6.0.5 has been validated for certain NIST SP 1086              configurations in terms of predicting the temperature NUREG/CR-4527, Volume 2           increase in an enclosure in accordance with NUREG-1824, Tatem et al. [2004]          Volume 5. In addition, NUREG/CR-4527, Volume 2 provides full scale test data of electrical panel fires in control room like structures. These tests are modeled using the CFAST, version 6.0.10.61027 and the results are documented in Report entitled "Evaluation of Main Control Room Abandonment Times at the Turkey Point Nuclear Plant." CFAST, version 6.0.10.61027 is found to provide a reasonable and conservative estimate of both the hot gas layer temperature and visibility as a function of time given the input fire size for a control room like enclosure. This information is documented in Appendix D of the Report entitled "Evaluation of Main Control Room Abandonment Times at the Turkey Point Nuclear Plant."
The MCR abandonment application falls within the non-dimensional parameter space for the NUREG-1824, Volume 5 V&V report as estimated using the methods described in NUREG-1934 [2011]. The application also falls within the model limits as specified in NIST SP 1026 [2009] and NIST SP 1041 [2008]. Additional V&V documentation is provided in NIST SP 1086 and Tatam et al. [2004] that expand the V&V parameter space from that included in NUREG-1824
[2007], including multiple compartment applications.
Generic Fire Modeling   Definition of zones of                    NUREG, Volume 5              Table J-2 provides a summary of the validation basis for the Treatments, Revision 0. influence about specific                     NIST SP 1026              empirical models used in the "Generic Fire Modeling classes of ignition sources.                 NIST SP 1041              Treatments" report.
Scenario screening for the                                             The "Generic Fire Modeling Treatments" report uses multi-compartment analysis.                  NIST SP 1086              CFAST, version 6.0.10.61027 in a simple geometry that Table J-2               minimizes the boundary heat losses given a volume. For the volume postulated, the configuration produces the most adverse result regardless of the actual dimensions used.
The application falls within the model limits as specified in NIST SP 1026 [2009] and NIST SP 1041 [2008]. Except for Revision 0                                                                                                                                           Page J-13

Florida Power& Light                                                                                                  Attachment J - Fire Modeling V&V Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation                  Application                        V & V Basis                                          Discussion the global equivalence ratio, the non-dimensional parameters fall within the V&V space of NUREG-1824 Volumes 1 and 5 [2007]. Although equivalence ratios are considered over a much larger range than addressed by the NUREG-1824, Volume 1 [2007] validation tests, the results are based on a single time point based on an equivalence ratio that is close to unity or lower and thus may fall directly within the NUREG-1824, Volume 1 [2007] V&V parameter space.
Additional validation results that consider higher the predictive capability under higher equivalence ratios are provided in NIST SP 1086 [2008).
Supplemental Generic      Definition of the ZOI about                    Table J-2              Table J-2 provides a summary of the validation basis for the Fire Model Treatments:    vented electrical panels.                 Mangs et al., 1994          empirical models used to generate the ZOIs in the "Generic Transient Ignition Source                                                                       Fire Modeling Treatments" report. This is the same Strength (Supplement 1).                                            Mangs et al., 1996          procedure and calculation used to generate the ZOls in Mangs et al., 2003          "Supplemental Generic Fire Model Treatments: Transient Mangs et al., 2006          Ignition Source Strength," Revision E.
NUREG/CR-4527, Volume 1, 1987       The predictive capability of the heat release rate model is NUREG/CR-4527, Volume 2,1988        validated against about thirty fire tests involving a wide range of panel sizes and ventilation configurations. The Melis et al., 2004         model is only capable of predicting the maximum heat release rate that can be supported within a robustly secured panel as defined in FAQ 08-0042 [NUREG/CR-6850, Supplement 1, 2010]. Fuels that may exhibit external burning (i.e., internal equivalence ratio greater than 1 - 1.3) are addressed by assuming an equivalence ratio of two, with half of the burning occurring inside the panel and half occurring outside the panel. This applies to PMMA, which is not a common material used in cable construction.
Page J-14 Revision 0                                                                                                                                           Page J-14

Florida Power & Light                                                                                                Attachment J - Fire Modeling V&V Table J-1 V & V Basis for Fire Models / Model Correlations Used in FPRA Calculation                  Application                        V & V Basis                                         Discussion Supplemental Generic      Definition of zones of                 NUREG-1824, Volume 5           The same methods developed in Generic Fire Modeling Fire Model Treatments:    influence about specific                    NIST SP 1026              Treatments, Revision 0 are used to generate additional hot Hot Gas Layer Tables      classes of ignition sources for             NIST SP 1041              gas layer tables and ZOI definitions. The treatment of (Supplement 2).           use in the FPRA.                                                      secondary applies to a two tray configuration located above Scenario screening for the                  NIST SP 1086              an electrical panel ignition source and used propagation multi-compartment analysis.          NUREG/CR-4527, Volume 2          rates recommended by NUREG/CR-6850 [2005] and Tatem et al. [2004]         validated for a wide range of cable compositions in Table J-2               NUREG/CR-7010 [2010].
NUREG-6850, 2005 NUREG/CR-7010, 2010 Supplemental Generic      Characterization of the heat                  Table J-2               The supplement provides an analysis of the transient fuel Fire Model Treatments:    release rate per unit area for                                        package source strength, which is used to determine the fire Transient Ignition Source transient ignition sources.                                           area and fire diameter. The revised fire diameters are used Strength (Supplement 3).                                                                        as input into the models used to derive the ZOls in the original "Generic Fire Modeling Treatments" report. The same calculations that were used in the original "Generic Fire Modeling Treatments" report are applied with new inputs.
Page J-15 Revision 00                                                                                                                                       Page J-15

Florida Power & Light                                                                                             Attachment J - Fire Modeling V&V Table J-2 V & V Basis for Fire Models I Model Correlations Used: Generic Treatment Correlations Location in the "Generic      OSubsequent Correlation  Fire Modeling      Original        Application          Original Correlation Range         Validation and       Limits in Treatments Treatments                                                                                 Verification Report" Flame Height  Page 18           Heskestad      Provides a limit                            \    1        Directly
[1981]         on the use of the  -5 Z log1 o                     Z,5    NUREG-1824, Zone of                                                   Volume 3 [2007]   41fiHt < 3000 Influence (ZOI)    In practice, wood and hydrocarbon      Indirectly fuels, momentum or buoyancy dominated, with diameters between      NUREG-1824, 0.05- 10 m (0.16 -33 ft).              Volume 5 [2007]
(Correlation used in CFAST)
Point Source  Page 19           Modak          Lateral extent of  Isotropic flame radiation. Compared    NUREG-1824,       Predicted heat flux at Model                           [1976]         ZOI -              with data for 0.37 m (1.2 ft) diameter Volume 3 [2007];  target is less than 5 comparison to      PMMA pool fire and a target located    SFPE [1999]       kW/m 2 (0.4 4 Btu/s-ft2) per other methods      at an L ratio of 10.                                     SFPE [1999].
R Method of     Page 19           Shokri et al. Lateral extent of  Pool aspect ratio less than 2.5.       SFPE [1999]       Ground based vertical Shokri and                      [1989]         ZOI -              Hydrocarbon fuel in pools with a                         target.
Beyler                                         comparison to      diameter between 1 - 30 m (3.3 - 98 other methods      ft).
Vertical target, ground level.
Method of     Page 20           Mudan          Lateral extent of  Round pools;                           SFPE [1999]       Total energy emitted by Mudan (and                      [1984]         ZOI -              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-16 Revision 00                                                                                                                                   Page J-16

Florida Power & Light                                                                                              Attachment J - Fire Modeling V&V Table J-2 V & V Basis for Fire Models I Model Correlations Used: Generic Treatment Correlations Location in the "Generic       Original                                                                 Subsequent Correlation   Fire Modeling     Reference        Application          Original Correlation Range          Validation and      Limits in Treatments Treatments                                                                                 Verification Report" Method of       Page 20          Shokri et al. Lateral extent of  Round pools;                           SFPE [1999]        Predicted heat flux at Shokri and                       [1989]         ZOl                Hydrocarbon fuel in pools with a                          target is greater than 5 Beyler                                                             diameter between 1 - 50 m (3.3 -       NUREG-1824,        kW/m2 (0.44 Btu/s-ft2 ) per 164 ft).                              Volume 3 [2007]    SFPE [1999].
Shown to produce most conservative heat flux over range of scenarios considered among all methods considered.
Plume heat      Page 22          Wakamutsu      Vertical extent of Fires with an aspect ratio of about 1  Wakamatsu et       Area source fires with fluxes                           et al. [2003]  ZOI                and having a plan area less than 1     al. [2003] (larger aspect ratio - 1. Used m2 (0.09 ft').                         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
[2008]             constraint in the fire model analysis for the cases evaluated.
Plume           Page 23          Yokoi          Vertical extent of Alcohol lamp assumed to effectively    NUREG-1824,        Area source fires with centerline                       [1960];        ZOI                be a fire with a diameter -0.1 m (0.33 Volume 3 [2007];   aspect ratio - 1. Used temperature                      Beyler                            ft).                                   SFPE Handbook      with plume flux
[1986]                                                                   of Fire Protection correlation; most severe Engineering,       of the two is used as Section 2-1        basis for the ZOI
[2008]             dimension.
Hydrocarbon     Page 51          SFPE           Determine heat     Hydrocarbon spill fires on concrete    None. Based on     None. Transition from spill fire size                  Handbook       release rate for   surfaces ranging from -1 to -10 m      limited number of  unconfined spill fire to of Fire        unconfined         (3.3 - 33 ft) in diameter.             observations.      deep pool burning Protection     hydrocarbon spill                                                            assumed to be abrupt.
Engineering,   fires.
Section 2-15 [2002]
Revision 0                                                                                                                                     Page J-17

Florida Power & Light                                                                                           Attachment J - Fire Modeling V&V Table J-2 V & V Basis for Fire Models I Model Correlations Used: Generic Treatment Correlations Location in the "Generic       Original                                                                Subsequent Correlation  Fire Modeling     Reference        Application          Original Correlation Range       Validation and      Limits in Treatments Treatments                                                                                Verification Report" Flame         Page 100          SFPE           Determine the      Corner fires ranging from -10 to     None. Based on     None. Offset is assumed extension                       Handbook       fire offset for    -1,000 kW (9.5- 948 Btu/s). Fires    limited number of  equal to the depth of the of Fire        open panel fires,  included gas burners and             observations,      ceiling jet from the Protection                        hydrocarbon pans.                                       experiments.
Engineering, Section 2-14 [2002]
Line source   Page 101          Delichatsios   Determine the      Theoretical development.             SFPE Handbook      None. Transition to area flame height                    [1984]         vertical extent of                                      of Fire Protection source assumed for the ZOI                                                 Engineering,       aspect plan ratios less Section 2-14       than four. Maximum of
[2008]             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       vertical extent of -1,000 kW (9.5 - 948 Btu/s). Fires   Correlation form of Fire        the ZOI            included gas burners and             is consistent with Protection                        hydrocarbon pans.                    other methods; Engineering,                                                           comparison to Section 2-                                                             dataset from 14 [2002]                                                              SFPE Handbook, Section 2-14
[2002] provides basis.
Air mass flow Page 140          Kawagoe        Compare            Small scale, % scale, and full scale Drysdale [1999];   None. SFPE [2004]
through                         [1958]         mechanical         single rooms with concrete and steel SFPE [2004]        spaces with a wide range opening                                        ventilation and    boundaries. Vent sizes and thus                         of opening factors.
natural            opening factor varied. Wood crib ventilation        fuels.
Revision 0                                                                                                                                   Page J-18 0

Florida Power& Light                                                                                                 Attachment J - Fire Modeling V&V Table J-2 V & V Basis for Fire Models / Model Correlations Used: Generic Treatment Correlations Location in the "Generic      OSubsequent Correlation   Fire Modeling      Original        Application          Original Correlation Range            Validation and     Limits in Treatments Treatments                                                                                   Verification Report" Line fire flame Page 210         Yuan et al. Provides a limit              Z                              None.             None.
height                           [1996]         on the use of the  0.002 <        <0.6                      Correlation form Zone of                                                     is consistent with Influence (ZOI);   In practice, from the base to severalmethods; times the flame height based on           omparimeto Extent of ZOI for  0.015 - 0.05 m (0.05 - 0.16 ft) wide     comparison to cable tray fires. gas burners.                             Yuan et al.
[1996] provides basis.
Cable heat      Page 210         Lee [1985]     Provides           Cables with heat release rates per       None.              Correlation predicts a release rate                                    assurance that     unit area ranging from about 100 -                          lower heat release rate per unit area                                   the method used    1,000 kW/m 2 (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              Z                             None.              None.
centerline                       [1996]         on the use of the  0.002 <        <0.6                      Correlation form temperature                                     Zone of            In practice, from the base to several    is consistent with Influence (ZOI);   Inpatcfo h aet eea times the flame height based on other methods; omparimeto Extent of ZOI for  0.015 - 0.05 m (0.05 - 0.16 ft) wide     comparison to cable tray fires. gas burners.                             Yuan et al.
[1996] provides basis.
Page J-19 Revision 00                                                                                                                                     Page J-19

Florida Power & Light                                                                                            Attachment J - Fire Modeling V&V Table J-2 V & V Basis for Fire Models / Model Correlations Used: Generic Treatment Correlations Location in the "Generic      Original                                                               Subsequent Correlation     Fire Modeling     Reference        Application           Original Correlation Range      Validation and     Limits in Treatments Treatments                                                                              Verification Report" Ventilation       Page 283         Babrauskas     Assessing the     Ventilation factors between 0.06 -    SFPE [2004]      None. Provides depth in limited fire size                  [1980]         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.
Page J-20 Revision 00                                                                                                                                   Page J-20

Florida Power & Light                                        Aftachment J - Fire Modeling V&V References
: 1)      ASTM E1355, "ASTM Standard Guide for Evaluating the Predictive Capability of Deterministic fire Models," American Society for Testing and Materials, West Conshohocken, PA, 2004.
: 2)      Babrauskas, V., "Estimating Room Flashover Potential," Fire Technology, No. 16, Vol. 2, pp. 94-104, 1980.
: 3)      Beyler, C. L., "Fire Plumes and Ceiling Jets," Fire Safety Journal, Volume 11, Number 1, Elsevier Sequoia S. A., Lausanne, Switzerland, 1986.
: 4)      Delicatsios, M. A., "Flame Heights in Turbulent Wall Fires with Significant Flame Radiation," Combustion Science andTechnology, 39, pp. 195-214. 1984.
: 5)      Drysdale, D. D., Fire Dynamics, 2 nd Edition, John Wiley and Sons, New York, New York, 1999.
: 6)      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.
: 7)      Heskestad. G., "Engineering Relations for Fire Plumes," Fire Safety Journal, 7:25-32, 1984.
: 8)      Hughes Associates, "Supplemental Generic Fire Model Treatments: Hot Gas Layer Tables," Supplement 2, Revision G, Baltimore, MD, October 10, 2011.
: 9)      Hughes Associates, "Supplemental Generic Fire Model Treatments: Transient Ignition Source Strength," Supplement 3, Revision E, Baltimore, MD, November 7,2011.
: 10)     Hughes Associates, "Supplemental Generic Fire Model Treatments: Closed Electrical Panels," Supplement 1, Revision B, Baltimore, MD, October 11, 2011.
: 11)     Hughes Associates, "Supplemental Generic Fire Model Treatments: Transient Target Response to Transient Ignition Source Fire Exposures," Supplement 4, Revision A, Baltimore, MD, October 13, 2011.
: 12)     Hughes Associates, Generic Fire Modeling Treatments, Project Number 1SPH02902.030, Revision 0, January 15, 2008.
: 13)     Hunt, S. P., "Maximum Fire Size in Closed Vented Electrical Panels," Fire Protection Information Forum, September, 2009.
: 14)     EPRI, "Evaluation of Heat Release Rates in Vertical Cabinet Fires," Electric Power Research Institute, Palo Alto, CA, 2011.
: 15)     Kawagoe, K., "Fire Behaviour in Rooms," Report Number 27, Building Research Institute, Tokyo, Japan, 1958.
: 16)     Lee, B. T., "Heat Release Rate Characteristics of Some Combustible Fuel sources in Nuclear Power Plants," NBSIR 85-3196, NIST, Gaithersburg, MD, 1985.
: 17)     Mangs, J. and Keski-Rahkonen, 0., "Full Scale Fire Experiments on Electronic Cabinets," Technical Research Centre of Finland (VTT) Publications 186, Technical Research Centre of Finland, Espoo, 1994.
Page J-21 Revision 00 Revision                                                                            Page J-21

Florida Power & Light                                        Affachment J - Fire Modeling V&V
: 18)     Mangs, J. and Keski-Rahkonen, 0., "Full Scale Fire Experiments on Electronic Cabinets I1," VTT Publications 269, Technical Research Centre of Finland, Espoo, 1996.
: 19)     Mangs, J., Paananen, J., and Keski-Rahkonen, 0., "Calorimetric Fire Experiments on Electronic Cabinets," Fire Safety Journal,38, pp. 165 - 186, 2003.
: 20)     Melis, S., Rigollet, L., Such, J.M., Casselman, C., "Modelling of Electrical Cabinet Fires Based on the CARMELA Experimental Program," Eurosafe Forum, 2004.
: 21)     Modak, A. T., "Thermal Radiation from Pool Fires," Technical Report 22361-5 RC-B-67, Factory Mutual Research, Norwood, MA, August, 1976.
: 22)     Mudan, K. S., "Thermal Radiation Hazards from Hydrocarbon Pool Fires," Prog.
Energy Combustion Sci., Vol. 10, 1984.
: 23)     NIST Special Publication 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.
: 24)     NIST Special Publication 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.
: 25)     NIST Special Publication 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.
: 26)     NUREG-0700, "Human-System Interface Design Review Guidelines,"
NUREG-0700, Rev. 2, Nuclear Regulatory Commission, Washington, DC, 2002.
: 27)     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.
: 28)     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.
: 29)     NUREG-1824, Volume 1, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications Volume 1: Main Report," NUREG-1 824 / 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.
: 30)     NUREG-1824, Volume 5, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications Volume 5: Consolidated Fire Growth and Transport Model," NUREG 1824/EPRI 1011999, Salley, M. H. and Kassawara, Revision 0                                                                          Page J-22

Florida Power & Light                                        Attachment J - Fire Modeling V&V R. P., NUREG-1824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., May, 2007.
: 31)     NUREG-1934, "Nuclear Power Plant Fire Modeling Applications Guide (NPP FIRE MAG)," Draft Report for Public Comment, NUREG-1934 / EPRI 1023259, Nuclear Regulatory Commission, Rockville, MD, 2011.
: 32)     NUREG/CR-4527, Volume 1, "An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets Part I: Cabinet Effects Tests,"
NUREG/CR-4527 / SAND86-0336, Volume 1, Chavez, J. M., Nuclear Regulatory Commission, Washington, DC, 1987.
: 33)     NUREG/CR-4527, Volume 2, "An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets Part II: Room Effects Tests,"
NUREG/CR-4527 / SAND86-0336, Volume 2, Chavez, J. M. and Nowlen, S. P.,
Nuclear Regulatory Commission, Washington, DC, November, 1988.
: 34)     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.
: 35)     NUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE) Volume 1: Horizontal Trays," Draft Report for Comment, McGrattan, K., Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, October, 2010.
: 36)     NUREG-6850 Supplement 1 (2010), "Fire Probabilistic Risk Assessment Methods Enhancements," Electric Power Research Institute (EPRI) 1019259, NUREG/CR-6850 Supplement 1, Nuclear Regulatory Commission, Washington, DC, September, 2010.
: 37)     Report 0020-0010-000, "Evaluation of Main Control Room Abandonment Times at the Turkey Point Nuclear Plant," Revision A, Hughes Associates, Inc.,
Baltimore, MD, May 14, 2012.
: 38)     SFPE Handbook of Fire Protection Engineering, Section 2-1, "Fire Plumes, Flame Height, and Air Entrainment," Heskestad, G., The SFPE Handbook of Fire Protection Engineering, 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 39)     SFPE Handbook of Fire Protection Engineering, Section 2-5, "Effect of Combustion Conditions on Species Production," Gottuk, D. and Lattimer, B. Y.,
The SFPE Handbook of Fire Protection Engineering, 4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 40)     SFPE Handbook of Fire Protection Engineering, Section 2-6, "Toxicity Assessment of Combustion Products," Purser, D. A., The SFPE Handbook of Fire Protection Engineering, 3 rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
: 41)     SFPE Handbook of Fire Protection Engineering, Section 2-14, "Heat Fluxes from Fires to Surfaces," Lattimer, B. Y., The SFPE Handbook of Fire Protection Engineering,4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
Revision 0                                                                          Page J-23

Florida Power & Light                                         Aftachment J - Fire Modeling V&V
: 42)     SFPE Handbook of Fire Protection Engineering, Section 2-14, "Heat Fluxes from Fires to Surfaces," Lattimer, B. Y., The SFPE Handbook of Fire Protection 0
Engineering,3 rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
: 43)     SFPE Handbook of Fire Protection Engineering, Section 2-15, "Liquid Fuel Fires," Gottuk, D. and White, D., The SFPE Handbook of Fire Protection Engineering,3 rd Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2002.
: 44)     SFPE Handbook of Fire Protection Engineering, Section 3-1, "Heat Release Rates," Babrauskas, V., The SFPE Handbook of Fire Protection Engineering,4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 45)     SFPE Handbook of Fire Protection Engineering, Section 3-4, "Generation of Heat and Gaseous, Liquid, and Solid Products in Fires," Tewarson, A., The SFPE Handbook of Fire Protection Engineering,4 th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
: 46)     Shokri, M and Beyler, C. L., "Radiation from Large Pool Fires," SFPE Journal of Fire Protection Engineering, Vol. 1, No. 4, pp. 141-150, 1989.
: 47)     Society of Fire Protection Engineers, "The SFPE Engineering Guide for Assessing Flame Radiation to External Targets from Pool Fires," Society of Fire Protection Engineers, National Fire Protection Association, Quincy, Mass., June, 1999.
: 48)     Society of Fire Protection Engineers, "Fire Exposures to Structural Elements,"
Engineering Guide, National Fire Protection Association, Quincy, Mass., May, 2004.
: 49)     Tatem, P.A., Budnick, E.K., Hunt, S.P., Trelles, J., Scheffey, J.L., White, D.A.,
Bailey, J., Hoover, J., Williams, F.W., "Verification and Validation Final Report for Fire and Smoke Spread Modeling and Simulation Support of T-AKE Test and Evaluation," NRL/MR/6180-04-8746, Naval Research Laboratory, Washington, D.C., 2004.
: 50)     Wakamatsu, T., Hasemi, Y., Kagiya, K., and Kamikawa, D., "Heating Mechanism of Unprotected Steel Beam Installed Beneath Ceiling and Exposed to a Localized Fire: Verification Using the Real-scale Experiment and Effects of the Smoke Layer," Proceedings of the Seventh International Symposium on Fire Safety Science, International Association for Fire Safety Science, London, UK, 2003.
: 51)     Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current," Report Number 34, Building Research Institute, Tokyo, Japan, 1960.
: 52)     Yuan, L. and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal, 27, 1996.
Page J-24 Revision 00                                                                          Page J-24

Florida Power & Light               Attachment K - Existing Licensing Action Transition K. Existing Licensing Action Transition 32 Pages Attached Page K-I Revision 00                                                                   Page K-1

Attachment K - Existing Licensing Action Transition Licensing Action                LA-01-19840327, Appendix R Exemption from Fixed Fire Suppression in the Control Room (III.G.3 Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section III.G.3 that requires installation of a fixed fire suppression system in the area, for which an alternate shutdown capability independent of cables, systems or components in that area, is provided. Specifically, FPL requested exemption from providing a fixed fire suppression system in the Control Room (Fire Area MM). This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-1 0-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations 4
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                         Run: 06/27/2012 14:41        Page: I of 32 0                                                                                            is

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-02-19840327, Appendix R Exemption from Having an Oil Collection Tank Sized to Contain the Entire Lube Oil System Inventory (111.0 Criteria)
Required Post-Transition         Yes Licensing Basis                  Exemption request per the 5/12/83 FPL submittal letter to the NRC provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section 111.0 for not having an oil collection tank sized to contain the entire lube oil system inventory. This exemption was granted by the NRC in an SER dated 3/27/84. The basis of the exemption includes:
                                 " Each unit has three reactor coolant pumps with an oil collection system which drains to a vented closed collection tank.
                                 " The quantity of lubricating oil in each pump is 200 gallons and the capacity of the oil collection tank is 265 gallons.
                                 " The components have been designed to be capable of withstanding a safe shutdown earthquake (SSE).
                                 " The collection tank is arranged such that if a failure of more than one RCP motor lube system occurred, the oil collection tank would overflow onto the lower containment floor.
                                 "The lubricating oil used in the RCP motors has a flashpoint of approximately 450&deg;F.
                                 "There are no ignition sources at the floor level of the lower containment.
In conclusion, the bases for previous NRC acceptance are still valid as substantiated by Attachment 2, Item LA-02-19840327 of PTN-FPER-08-007.
Fire Areas                       ID                     Description P (Unit 4)             Unit 4 Containment Building Q (Unit 3)             Unit 3 Containment Building References                       Document ID 1983-05-12 [Letter] - FPL-NRC, L-83-295, 10 CFR 50 Appendix R 111.0 - Oil Collection System Evaluation The FPL submittal letter to the NRC stated:
                                 "Discussions with the NRC staff have recently indicated that there may be some concern regarding compliance of the Turkey Point Units 3 and 4 Reactor Coolant Pump Lubricating Oil Collection System with the requirements of 10 CFR 50 Appendix R Section 111.0 with specific regard to system capacity.
The RCP Lube Oil Collection System was designed and installed as proposed in our letters L-80-140 dated May 7, 1980 and L-80-336 dated October 7, 1980 and approved in NRC letter dated December 8, 1980. The system capacity, as designed, was based on retaining the entire lube oil system inventory of one pump and an additional amount of miscellaneous leakage from all other pumps. We believe this to be a reasonable design assumption based on the low potential for simultaneous failure of independent lube oil systems.
Recent discussions held with NRC staff have indicated that this interpretation may be inconsistent with current thinking and that an exemption request should be considered for this design. We therefore, submit in accordance with 10 CFR 50.12, the following request for exemption and supporting justification:
An exemption is requested for Turkey Point Units 3 and 4 from that portion, of 10 CFR 50, Appendix R, Section 111.0, which requires the reactor coolant pump lube oil collection system to be capable of containing the entire lube oil system inventory.
The installed lube oil collection system is capable of containing the entire inventory of one reactor coolant pump lube oil system (200 gallons) with sufficient excess capacity to contain any potential miscellaneous leakage from the remaining two pumps (65 gallons). The oil collection tank is located within the containment building
[sic, recte Containment Building], outside of the biological shield wall at the base mat elevation. The tank is situated away from sources of ignition such as hot reactor coolant system piping and is located as far as practicable away from safe shutdown equipment and cable.
Each reactor coolant pump motor is provided with low level alarms on the upper and lower oil reservoir such that a decrease of 1" (approximately 5 gallons) of level in either reservoir will actuate an alarm in the control room. This alarm and existing operating procedures assure operator investigation at the first indication of leakage.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                         Run: 06/27/2012 14:41         Page: 2 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-02-19840327, Appendix R Exemption from Having an Oil Collection Tank Sized to Contain the Entire Lube Oil System Inventory (111.0 Criteria)
Evaluation The lubricating oil used in the reactor coolant pump motor has a flash point of over 450'F providing additional assurance that a fire would not occur.
The lube oil collection system is seismically designed and will remain functional during a seismic event. Additionally, on the basis of historical or statistical seismic activity, Turkey Point is located in a seismically inactive area, far from any recorded damaging shocks.
In conclusion, the existing reactor coolant pump oil collection system provides adequate capability for collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems. Additional modifications would not significantly augment or materially enhance the safety of the plant since they would not aid in the prevention of fire damage to redundant components essential for safe shutdown. We therefore conclude that this is an acceptable exemption to Appendix R to 10 CFR Part 50, Section 111.0."
Document ID 1984-03-27 [SER Sections 12.2, 12.3, and 12.4] - NRC-FPL, Exemption Requests for Turkey Point Plant Units 3 and 4 Evaluation The NRC Safety Evaluation which granted this exemption stated:
                                 "12.2 Discussion Each unit has three reactor coolant pumps with an oil collection system which drains to a vented closed collection tank. The quantity of lubricating oil in each pump is 200 gallons and the capacity of the oil collection tank is 265 gallons. The components have been designed to be capable of withstanding a safe shutdown earthquake (SSE). The collection tank is arranged such that if a failure of more than one RCP motor lube system occurred, the oil collection tank would overflow onto the lower containment floor. The lubricating oil used in the RCP motors has a flashpoint of approximately 450&deg;F. There are no ignition sources at the floor level of the lower containment.
12.3 Evaluation The RCP motor lube oil system does not comply with Section 111.0 because the oil collection tank is not sized to contain the entire lube oil system inventory. The oil collection tank is provided with sufficient capacity to hold the total lube oil inventory of one reactor coolant pump with margin and is designed so that any overflow will be drained to a safe area. We agree with the licensee that this combination of features is acceptable.
12.4 Conclusion Based on the above evaluation, the existing RCP motor lube oil collection system provides a level of safety equivalent to the technical requirements of Section 111.0 and, therefore, the exemption should be granted."
Document ID PTN-FPER-08-007 Rev. 2 [Transitioning LA Bases Review &Verification] - NFPA 805 Transition Review of Existing Licensing Actions Evaluation Attachment 2 of PTN-FPER-08-007 documents verification of configurations discussed in the above Licensing Action.
Chapter 3 References             Reference 3.3.12 Reactor Coolant Pumps Fire Safety Analysis Data Manager (4.1) 0                                                                                             Turkey Point 0
Run: 06/27/2012 14:41 0
Page: 3 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-03-19840327, Appendix R Exemption from Providing 3-Hour Rated Fire Barriers Between Fire Zones 11, 12, 13, 14, 15 and 16 (III.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIl.G.2.a that requires separation of cables and components in redundant trains used for safe shutdown. Specifically, by FPL letters L-83-225 and L-83-347, FPL requested exemption from providing three-hour fire barriers between the RHR heat exchangers, the RHR pumps, and for the power cables for the 3A RHR pump. This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                       Run: 06/27/2012 14:41      Page: 4 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-04-19840327, Appendix R Exemption from Providing a 3-Hour Rated Fire Door for the Charging Pump Rooms on the West Walls of Fire Areas N & 0 (llI.G.2.a Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIl.G.2.a that requires three-hour rated barriers be provided to separate redundant trains of equipment. Specifically, by FPL letters L-83-225 and L-83-347, FPL requested exemption from providing a three-hour rated fire door on the west wall of Fire Area N and Fire Area 0 in place of the existing radiation shield door. This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because according to FPL engineering evaluation PTN-FPER-1 1-003, the open passageway is adequate for the hazard. No credit is taken at all for the shield door.
References                      Document ID PTN-FPER-11-003 Rev. 0 - Turkey Point GL 86-10 Evaluation of Adequacy of Radiation Shield Doors in Charging Pump Rooms for Preventing Spread of Fire Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                          Run: 06/27/2012 14:41     Page: 5 of 32 0                                                                                           0                                                                                           0

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-05-19840327, Appendix R Exemption from Providing a Fire Suppression System for the Charging Pump Rooms Fire Areas N and 0 (IIl.G.2.c Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section ill.G.2.c that requires installation of a fixed fire suppression system in the area, for which redundant trains of safe shutdown equipment are not separated by a three-hour rated fire barrier. Specifically, by FPL letters L-83-225 and L-83-347, FPL requested exemption from providing a fixed fire suppression system in the Unit 3 and Unit 4 charging pump rooms (Fire Areas 0 and N, respectively). This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-1 0-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-10-010 Rev. 3 -Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                        Run: 06/27/2012 14:41        Page: 6 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-06-19840327, Appendix R Exemption from Providing Emergency Lighting Units in the Units 3 & 4 Containment to Facilitate the Manual Operation of Accumulator Stop Valves and RHR Inlet Isolation Valves (lll.J Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section Illi.J, which requires emergency lighting units with at least an 8 hour battery power supply to be provided in all areas needed for operation of safe shutdown equipment and in access and egress routes thereto.
Specifically, FPL requested exemption from providing 8 hour battery backed lights in the unit 3 and unit 4 Containments (Fire Areas Q and P respectively). This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because NFPA 805 does not require 8 hour battery backed emergency lights.
References                       Document ID NEI 04-02 Rev. 2 - Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48 (c)
Fire Safety Analysis Data Manager (4.1) 0                                                                                           Turkey Point 0
Run: 06/27/2012 14:41 0
Page: 7 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria)
Required Post-Transition         Yes Licensing Basis                  Exemption request per the 4/8/83 FPL submittal letter, as supplemented by the 6/6/83 and 9/30/83 letters to the NRC, provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R IIl.G.2.f that requires separation of cables and equipment and associated non-safety circuits of redundant trains by a non
                                 -combustible radiant energy shield where separation distance is less than 20'-0". Specifically, exemptions were requested from providing non-combustible radiant energy shields between Pressurizer Backup Heaters Group A and B at point of termination at the base of the pressurizer, Power Operated Relief Valves PCV-3/4-455C and PCV-3/4-456, Block Valves MOV-3/4-536 and MOV-3/4-535, and Pressurizer Level Transmitters LT-3/4-459 and LT-3/4-460. This exemption was granted by the NRC in an SER dated 3/27/84. The basis of the exemption includes:
* Normal access is through the personnel airlock, access for fire fighting is also available through the emergency air lock and the access hatch.
                                 " No hose stations are located inside Containment; however, four portable fire extinguishers are available inside Containment.
* Ionization type fire detectors are installed in selected locations throughout containment.
                                 " This area within the pressurizer shield walls is void of in situ combustibles and is inaccessible during plant operation.
                                 ' All cabling inside the missile shield walls are routed in conduit.
                                 " Safe shutdown cable trays are generally located outside the secondary shield wall and run from about 5 ft. to 65 ft. above the floor.
                                 " Barrier separation of redundant trains is not provided, however, all cables are covered with a fire retardant mastic material.
                                 ' Redundant cable trays are separated from each other by minimum horizontal and vertical distances of 1 1/2 ft. and 4 ft., respectively.
                                 ' Combustible material in the area consists of lubricating oil contained in various components, and fire retardant coated cables.
                                 ' Access into Containment is controlled and limited with the possibility of introducing transient combustibles substantially reduced.
                                 ' FPL had proposed to either reroute redundant cables and equipment to provide 20 feet of separation, free of intervening combustibles, or provide 1-hour fire rated barriers and radiant energy shields on all redundant cables in both units except for the redundant equipment associated with operation of the pressurizer. The modifications FPL proposed included the following:
o Reroute the control cables for AOV-3-460 through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valves AOV                                       200A, B, and C.
o Reroute the control cables for AOV-3-31 OA through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valve AOV                                       31OB.
o Provide a minimum of 20 feet of separation for cables between two trains of reactor coolant system hot and cold leg temperature instrumentation where physically possible. Route the two trains, when provided, through separate penetration areas.
o Provide 1-hour rated protection on the conduit for LT-3-460 to the maximum extent possible in the pressurizer missile shield wall area where separation from conduits for LT-3-459 is less than 20 feet.
o Provide dedicated portable emergency lighting units outside the Containment for Containment entry to facilitate manual operation of valves.
o Reroute the control cables for valve AOV-4-460 through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valves AOV-4-200 A, B, and C as far as physically possible.
o Reroute the control cables for valve AOV-4-387 through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valve HCV 137 as far as physically possible.
o Reroute the control cables for AOV-4-310A through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valve AOV                                      310B as far as physically possible.
o Provide 1-hour rated protection to the conduit for LT-4-460 to the maximum extent possible in the pressurizer missile shield wall area where separation from conduits for LT-4-459 is less than 20 feet.
o FPL will install radiant energy shields to separate the charging line isolation valve and associated cabling.
In conclusion, the bases for previous NRC acceptance are still valid as substantiated by Attachment 2, Item LA-07-19840327 of PTN-FPER-08-007 except as clarified in Attachment T of this License Amendment Request.
Fire Safety Analysis Data Manager (4.1)                                                              Turkey Point                                                      Run: 06/27/2012 14:41     Page: 8 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (Ill.G.2.f Criteria)
Fire Areas                      ID                      Description P (Unit 4)              Unit 4 Containment Building Q (Unit 3)              Unit 3 Containment Building References                      Document ID 1983-04-08 Rev. 0 [Letter] - FPL-NRC, L-83-225, 10 CFR 50 App R Revision of Submitted Information Evaluation This FPL letter to the NRC provided the original Rev. 0 Fire Protection Report submittal including exemptions. This Rev. 0 submittal, however, is not attached to the letter. There is a note stating to refer to L-83-347 for the attachment. The L-83-347 letter provides the complete Rev. 1 Fire Protection Report including exemptions.
This is the reference that is being used as the original submittal for Turkey Point since the Rev. 0 Fire Protection Report is not retrievable and the NRC provided the safety evaluation based on the Rev. 1 Fire Protection Report submittal.
Document ID 1983-06-06 Rev. 1 [Fire Area P & Q, Exemption Request P-1 and Q-1] - FPL-NRC, L-83-347, 10 CFR 50 Appendix R Fire Protection Evaluation The FPL submittal to the NRC for Fire Area P Exemption P-1 stated:
                                "C. PROPOSED MODIFICATIONS
: 1. Reroute the control cables for valve AOV-4-460 through West Penetration Area and maintain the separation in excess of 20' from the cables for valves AOV                                        200A, B and C as far as physically possible.
: 2. Reroute the control cables for valve AOV-4-387 through West Penetration Area and maintain the separation in excess of 20' from the cables for valve HCV-4-137 as far as physically possible.
: 3. Reroute the control cables for AOV-4-31 OA through West Penetration Area and maintain the separation in excess of 20' from the cables for valve AOV-4-31 OB as far as physically possible.
: 4. Provide a minimum of 20' of separation for cables between two trains of Reactor Coolant System hot and cold leg temperature instrumentation, wherever physically possible. Route the two trains, when provided, through separate penetration areas.
: 5. Provide one-hour rated protection to the conduit for LT-4-460 to the maximum extent possible in the pressurizer missile shield wall area, where separation from conduits for LT-4-459 is less than 20'.
: 6. Provide dedicated portable emergency lighting outside containment units for Containment entry to facilitate manual operation of the valves.
D. EXEMPTION REQUEST
: 1. FPL requests exemption from that portion of Appendix R to 10CFR [sic, recte 10 CFR] Part 50 Section III G.2.f [sic, recte III.G.2.f] that requires separation of cables and equipment and associated non-safety circuits of redundant trains by a non-combustible radiant energy shield where separation distance is less than 20'-0".
Specifically exemptions are requested from providing non-combustible radiant energy shields between Excess Letdown Isolation Valves AOV-4-387 and HCV-4-137, Pressurizer Backup Heaters Group A and B at point of termination at the base of the pressurizer, Charging Line Isolation Valves AOV-4-310A and AOV-4-310B, Power Operated Relief Valves PCV-4-455C and PCV-4-456, Block Valves MOV-4-536 and MOV-4-535, and Pressurizer Level Transmitters LT-4-459 and LT-4-460."
                                "E. JUSTIFICATION FOR EXEMPTION Fire Safety'Analysis Data Manager (4.1)                                                            Turkey Point                                                          Run: 06/27/2012 14:41      Page: 9 of 32 0                                                                                             0                                                                                             0

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (Ill.G.2.f Criteria)
Evaluation Fire Area P is the Unit 4 Containment Building. Accessibility in the containment [sic, recte Containment] is available through the personnel air lock and the emergency air lock during operation. However, this accessibility is controlled administratively. Insitu [sic, recte In situ] combustible sources in this fire area are lubricating oil contained in Reactor Coolant Pumps, Manipulator Crane Trolley, CRDM Coolers, Normal Containment Coolers, Emergency Containment Coolers, Hydraulic Snubbers, Sump Pumps,Containment Polar Crane, Reactor Coolant Drain Tank Pumps, and electrical cables routed in trays.
The largest potential combustible source in containment [sic, recte Containment] is the lube oil associated with the Reactor Coolant Pumps. This consists of approximately 200 gallons per pump. The Reactor Coolant Pumps have an oil collection system as required by Section 111.0 of Appendix R, thus precluding this potential combustible source from initiating a fire in Containment. The second largest combustible source in Containment is the lubricating oil associated with the Containment Polar Crane and the Manipulator Crane trolleys. Since the Manipulator Crane and Containment Polar Crane are stored during plant operation, the potential for an oil spill from this equipment is considered extremely unlikely. The remaining liquid combustibles in Containment consist of the small amounts of lubricating oil associated with the specific equipment identified in the above list. These potential combustible sources are localized, generally well separated, and confined within the steel structure of the specific equipment. The lubricating oil systems associated with the equipment are not pressurized, and therefore the potential for oil spray is unlikely.
All the cables in cable trays are sprayed with Flamastic 71A or 77. This fire retardant coating, although not a valid fire barrier, acts as a radiant energy shield to provide a substantial level of protection against a postulated exposure fire. Further the location of the trays at elevations ranging from 18'-7" to 80' outside the secondary shield wall and configuration of containment building [sic, recte the Containment Building] itself limits the possibility of cable damage due to stratification of hot gases or other products of combustion.
Redundant equipment associated with operation of the pressurizer, including the pressurizer heaters, PORV's, Block Valves and Level Transmitters are located on or within the pressurizer missile shield wall as dictated by plant design criteria. This area is void of significant insitu [sic, recte in situ] combustibles, and during plant operation, considered inaccessible to transient combustibles. Cable to this equipment is routed in a well separated manner to the maximum extent possible and rapidly achieves 20' separation as it leaves the area of the pressurizer.
Charging line isolation valves are located within the biological shield wall in an area having no insitu [sic, recte in situ] combustibles in proximity to the valves. Cables to these redundant valves are routed in steel conduit in the area of the valves and following proposed modifications, will rapidly achieve 20' separation as they leave the area of the valves.
Excess Letdown Isolation Valves are located within the biological shield wall in an area having no insitu [sic, recte in situ] combustibles in proximity to the valves.
Cables to these redundant valves are routed in steel conduit in the area of the valves and following proposed modifications, will rapidly achieve 20' separation as they leave the area of the valves.
Fire protection features, existing and proposed for [sic, recte proposed, for] this fire zone are both active and passive in nature. Passive protection is provided for the redundant equipment and instrumentation and their associated cables by existing and proposed physical separation and one-hour rated protection. Active fire protection is provided by the existing fire detection systems in the electrical penetration area which will provide rapid early warning of fire before significant development occurs."
                                "In conclusion, it is FPL's position that the proposed and existing fire protection features for Fire Area P provide a level of fire protection consistent with the fire hazards, both insitu [sic, recte in situ] and transient, identified for this fire area and that these features provide a high level of assurance that at least one train of safe shutdown equipment and cables will remain free of fire damage."
The FPL submittal to the NRC for Fire Area Q Exemption Q-1 stated:
                                "C. PROPOSED MODIFICATIONS
: 1. Reroute the control cables for valve AOV-3-460 through the West Penetration Area and maintain the separation in excess of 20' from the cables for valves AOV-3-200A, B and C, as far as physically possible.
: 2. Reroute the control cables for AOV-3-310A through West Penetration Area and maintain the separation in excess of 20' from the cables for valve AOV-3-31 OB, as far as physically possible.
Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                             Run: 06/27/2012 14:41      Page: 10 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (llI.G.2.f Criteria)
Evaluation
: 3. Provide a minimum of 20' of separation for cables between two trains of Reactor Coolant System hot and cold leg temperature instrumentation, where physically possible. Route the two trains, when provided, through separate penetration areas.
: 4. Provide one-hour rated protection on the conduit for LT-3-460 to the maximum extent possible in the pressurizer missile shield wall area where separation from conduits for LT-3-459 is less than 20'.
: 5. Provide dedicated portable emergency lighting units outside Containment for Containment entry to facilitate manual operation of valves.
D. EXEMPTION REQUEST
: 1. FPL requests exemption from that portion of Section IIIG.2.f [sic, recte IIl.G.2.f] of Appendix R to 10CFR [sic, recte 10 CFR] Part 50 that requires separation of cables and equipment and associated non-safety circuits of redundant trains by a non-combustible radiant energy shield where separation distance is less than 20'-0".
Specifically exemptions are requested from providing non-combustible radiant energy shields between Excess Letdown Isolation Valves AOV-3-387 and HCV-3-137, Pressurizer Backup Heaters Group A and B at point of termination at the base of Pressurizer, Charging Line Isolation Valves AOV-3-31 OA and AOV-3-31 GB, Power Operated Relief Valves PCV-3-455C and PCV-3-456, Block Valves MOV-3-536 and MOV-3-535, and Pressurizer Level Transmitters LT-3-459 and LT-3-460."
                                "E. JUSTIFICATION FOR EXEMPTION Fire Area Q is the Unit 3 Containment Building. Accessibility in the Containment is available through the personnel air lock and the emergency air lock during operation. However, this accessibility is controlled administratively.
Insitu [sic, recte In situ] Combustible [sic, recte combustible] sources in this fire area are lubricating oil contained in Reactor Coolant Pumps, Maniuplator [sic, recte Manipulator] Crane Trolley, CRDM Coolers, Normal Containment Coolers, Emergency Containment Coolers, Hydraulic Snubbers, Sump Pumps, Containment Polar Crane, Reactor Coolant Drain Tank Pumps, and electrical cables routed in trays.
The largest potential combustible source in Containment is the lube oil associated with the Reactor Coolant Pumps. This consists of approximately 200 gallons per pump. The Reactor Coolant Pumps have an oil collection system as required by Section 111.0 of Appendix R, thus precluding this potential combustible source from initiating a fire in Containment. The second largest combustible source in Containment is the lubricating oil associated with the Containment Polar Crane and the Manipulator Crane trolleys. Since the Manipulator Crane and Containment Polar Crane are stored during plant operation, the potential for an oil spill from this equipment is considered extremely unlikely. The remaining liquid combustibles in Containment consist of the small amounts of lubricating oil associated with the specific equipment identified in the above list. These potential combustible sources are !ocalized, generally well separated, and confined within the steel structure of the specific equipment. The lubricating oil systems associated with the equipment are not pressurized, and therefore the potential for oil spray is unlikely.
All the cables in cable trays are sprayed with Flamastic 71A or 77. This fire retardant coating, although not a rated fire barrier, acts as a radiant energy shield to provide a substantial level of protection against postulated exposure fires. Further, the location of the trays at elevations ranging from 19'-5" to 80', outside of the secondary shield wall and the configuration of Containment [sic, recte of the Containment] Building itself precludes the possibility of cable damage due to stratification of hot gases or other products of combustion.
Redundant equipment associated with operation of the pressurizer, including the pressurizer heaters, PORV's, Block Valves and Level Transmitters are located on or within the pressurizer missile shield wall as dictated by plant design criteria. This area is void of significant insitu [sic, recte in situ] combustibles, and during plant operation, considered inaccessable [sic, recte inaccessible] to transient combustibles. Cable to this equipment is routed in a well separated manner to the maximum extent possible and rapidly achieves 20' separation as it leaves the area of the pressurizer.
Charging line isolation valves are located within the biological shield wall in an area having no insitu [sic, recte in situ] combustibles in proximity to the valves. Cables to these redundant valves are routed in steel conduit in the area of the valves and, following proposed modifications, will rapidly achieve 20' separation as they leave the area of the valves.
Excess Letdown Isolation Valves are located within the biological shield wall in an area having no insitu [sic, recte in situ] combustibles in proximity to the valves.
Cables to these redundant valves are routed in steel conduit in the area of the valves and, following proposed modifications, will rapidly achieve 20' separation as they 0
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                               Run: 06/27/2012 14:41 0 Page: 11 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria)
Evaluation leave the area of the valves.
Fire protection features, existing and proposed for this fire zone are both active and passive in nature. Passive protection is provided for the redundant equipment and instrumentation and their associated cables by existing and proposed physical separation and one-hour rated protection.
Active fire protection is provided by the existing fire detection system in the electrical penetration areas which will provide rapid early warning of fire before significant development occurs."
                                 "In conclusion, it is FPL's position that the proposed and existing fire protection features for the Fire Area Q provide a level of fire protection consistent with the fire hazards, both insitu [sic, recte in situ] and transient, identified for this fire area and that these features provide a high level of assurance that at least one train of safe shutdown equipment and cables will remain free of fire damage."
Document ID 1983-09-30 [Attachment] - FPL-NRC, L-83-503, 10 CFR 50 App R Fire Protection Supplemental Information Evaluation This FPL letter clarified and supplemented previous information. Item 2 addressed a radiant energy shield modification. Item 6 withdrew the commitment to provide radiant energy shields between the excess letdowns isolation valves AOV-4-387 and HCV-4-137. It was agreed that the exemption for Item 6 was not needed since the valves were much more than 20' apart. The letter stated:
                                 "2. A radiant energy shield will be located between the charging line isolation valves (4-31 0A and 4-31 OB) located in the Unit 4 containment [sic, recte Containment].
The radiant energy shield will be designed such that line of sight communication between these two valves will be prevented. Similar modifications for the Unit 3 charging line isolation valves will be provided if the need is identified during the future Unit 3 containment [sic, recte Containment] walkdown."
                                 "6. The exemptions requested from providing radiant energy shields between the excess letdowns isolation valves AOV-4-387 and HCV-4-137 is withdrawn. It was agreed that this exemption was not needed since the valves were much more than 20' apart."
Document ID 1984-03-27 [SER Sections 8.2, 8.3 and 8.4] - NRC-FPL, Exemption Requests for Turkey Point Plant Units 3 and 4 Evaluation The NRC Safety Evaluation stated the following:
                                "8.2     Discussion Fire Area P (Containment Building - Unit 4)
Discussion Fire Area Q (Containment Building - Unit 3)
Each Containment Building is essentially open and considered to be one fire area. There are three intermediate floor levels at the 14 ft., 30 ft., and 58 ft. elevations.
Normal access is through the personnel air lock, however, access for fire fighting is also available through the emergency air lock and the access hatch.
No hose stations are located inside containment [sic, recte Containment], however, four portable fire extinguishers are available inside containment [sic, recte Containment] for use by the fire brigade. Ionization type fire detectors are installed in selected locations throughout containment [sic, recte Containment]. No automatic suppression is provided inside containment [sic, recte Containment].
Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                           Run: 06/27/2012 14:41       Page: 12 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lIl.G.2.f Criteria)
Evaluation Redundant shutdown equipment inside containment [sic, recte Containment] includes both normal and emergency containment coolers, accumulator discharge valves, letdown isolation valves and excess letdown isolation valves, RHR inlet isolation valves, pressurizer heater and sprays, and various motor operator valves and instrumentation transmitters. Safe shutdown cable trays are generally located outside the secondary shield wall and run from about 5 ft. to 65 ft. above the floor.
Barrier separation of redundant trains is not provided, however, all cables .are covered with a fire retardant mastic material. Redundant cable trays are separated from each other by minimum horizontal and vertical distances of 1 1/2 ft. and 4 ft., respectively.
Combustible material in the area consists of lubricating oil contained in various components, and fire retardant coated cables. Since access into containment [sic, recte Containment] is controlled and limited, the possibility of introducing transient combustibles is substantially reduced.
The licensee proposes the following Unit 4 modifications:
: 1. Reroute the control cables for valve AOV-4-460 through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valves AOV-4-200 A, B and C as far as physically possible.
: 2. Reroute the control cables for valve AOV-4-387 through the West Penetation [sic, recte Penetration] Area and maintain the separation in excess of 20 feet from the cables for valve HCV-4-137 as far as physically possible.
: 3. Reroute the control cables for AOV-4-310A through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valve AOV                                         310B as far as physically possible.
: 4. Provide a minimum of 20 feet of separation for cables between two trains of reactor coolant system hot and cold leg temperature instrumentation, wherever physically possible. Route the two trains, when provided, through separate penetration areas.
: 5. Provide 1-hour rated protection to the conduit for LT-4-460 to the maximum extent possible in the pressurizer missile shield wall area where separation from conduits for LT-4-459 is less than 20 feet.
: 6. Provide dedicated portable emergency lighting outside the containment unit for containment [sic, recte Containment] entry to facilitate manual operation of the valves.
: 7. The licensee will install radiant energy shields to separate the charging line isolation valve and associated cabling.
The licensee proposes the following Unit 3 modifications:
: 1. Reroute the control cables for AOV-3-460 through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valves AOV                                         200A, B, and C.
: 2. Reroute the control cables for AOV-3-310A through the West Penetration Area and maintain the separation in excess of 20 feet from the cables for valve AOV                                         310B.
: 3. Provide a minimum of 20 feet of separation for cables between two trains of reactor coolant system hot and cold leg temperature instrumentation where physically possible. Route the two trains, when provided, through separate penetration areas.
: 4. Provide 1-hour rated protection on the conduit for LT-3-460 to the maximum extent possible in the pressurizer missile shield wall area where separation from conduits for LT-3-459 is less than 20 feet.
: 5. Provide dedicated portable emergency lighting units outside the containment [sic, recte Containment] for containment [sic, recte Containment] entry to facilitate manual operation of valves.
: 6. The licensee will install radiant energy shields to separate the charging line, isolation valves and associated cabling.
8.3 Evaluation The technical requirements of Section III.G are not met because radiant energy barriers are not provided for redundant trains of safe shutdown equipment and cabling is separated by less than 20 feet.
Section III.G calls for installation of a radiant energy shield between redundant safe shutdown equipment and cables that are closer to each other than 20 feet 0
Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                      Run: 06/27/2012 14:41 0Page: 13 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-07-19840327, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lIl.G.2.f Criteria)
Evaluation horizontal distance. The licensee has proposed to either reroute redundant cables and equipment to provide 20 feet of separation, free of intervening combustibles, or provide 1-hour fire rated barriers and radiant energy shields on all redundant cables in both units except for the redundant equipment associated with operation of the pressurizer, including the pressurizer heaters, PORVs, block valves and level transmitters located on or within the pressurizer missile shield walls. These areas are void of in situ combustibles and are inaccessible during plant operation. All cabling inside the missile shield walls are routed in conduit. Due to the cable and equipment configuration and location within the containment [sic, recte Containment] and the restricted access of these sub-areas during plant operations, an exposure fire involving the accumulation of significant quantities of transient combustible materials is unlikely. Because there are only a few cables in these sub-areas and all cables are routed in conduit, a fire of sufficient magnitude to damage redundant cables or components is also unlikely.
8.4 Conclusion Based on the above evaluation, the existing protection for the containment areas provides a level of fire protection equivalent to technical requirements of Section III.G of Appendix R. Therefore, exemption should be granted."
Document ID PTN-FPER-08-007 Rev. 2 [Transitioning LA Bases Review & Verification] - NFPA 805 Transition Review of Existing Licensing Actions Evaluation Attachment 2 of PTN-FPER-08-007 documents verification of configurations discussed in the above Licensing Action.
Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                         Run: 06/27/2012 14:41      Page: 14 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-08-19840327, Appendix R Exemption from Providing the Enclosure of One Charging Pump in a 1-Hour Barrier for Fire Areas N and 0 (lll.G.2.c Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIl.G.2.c that requires the enclosure of cable and equipment of one redundant train in a fire barrier having a one-hour rating. Specifically, FPL requested exemption from the enclosure of one charging pump in a one-hour rated fire barrier in the unit 3 and unit 4 charging pump rooms (Fire Areas 0 and N respectively). This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-10-010 Rev. 3 -Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                     Run: 06/27/2012 14:41     Page: 15 of 32
      .0                                                                                           0 0

0                                                                                           *0 Attachment K - Existing Licensing Action Transition Licensing Action                 LA-09-19840327, Appendix R Exemption from Total Enclosure of One Diesel Generator Radiator Room By a Three-Hour Rated Fire Barrier (llI.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section Ili.G.2.a that requires separation of cables and components in redundant trains by a fire barrier having a 3-hour rating. Specifically, exemption was requested from total enclosure of one Diesel Generator Radiator Room by three-hour rated fire barrier. This exemption was granted by the NRC in an SER dated 3/27/84.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                         Turkey Point                                                        Run: 06/27/2012 14:41     Page: 16 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-1 0-19870812, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (lll.G.2.d Criteria)
Required Post-Transition        Yes Licensing Basis                 Exemption request per the 04/25/86 FPL submittal letter, as supplemented by the 2/11/1987 letter to the NRC, provides justification for the exemption from the provisions of Section IIl.G.2.d of Appendix R to 10CFR50 that requires 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. Specifically, exemption was requested for intervening combustibles inside the primary containment for each of Turkey Point Units 3 and 4 (Fire Areas Q and P, respectively). This exemption was granted by the NRC in an SER dated 8/12/87.
The basis of the exemption includes:
                                " Reactor Coolant Pump Motors are located in separate cubicles and are fitted with oil collection assemblies, other oil sources are located away from safe shutdown cables.
                                " The major combustible materials are lubricating oil and cable insulation.
                                " To the maximum extent possible cable is rerouted to achieve at least 20-feet clear separation.
                                " Controlled personnel access to Containment reduces the potential for significant quantities of transient combustibles.
                                 " The cables in cable trays are either coated with a fire retardant coating or qualified to the requirements of IEEE Standard 383-1974.
                                " The probability of large amounts of transient combustibles being accumulated is low.
                                 " Portable fire extinguishers are located inside Containment and in the immediate vicinity of each personnel access hatch.
                                 " Ionization type smoke detectors are installed in the electrical penetration areas which alarm in the Control Room.
                                 " The large volume of the Containment.
In conclusion, the bases for previous NRC acceptance are still valid as substantiated by Attachment 2, Item LA-1 0-19870812 of PTN-FPER-08-007 except as clarified in Attachment T of this License Amendment Request.
Fire Areas                      ID                     Description P (Unit 4)             Unit 4 Containment Building Q (Unit 3)             Unit 3 Containment Building References                      Document ID 1986-04-25 [Attachment 4] - FPL-NRC L-86-164, 10 CFR 50 Appendix R Fire Protection Evaluation The FPL submittal letter provides the details of the exemption request made to the NRC. It states the following:
                                "Exemption Request Pursuant to 10CFR50.12, FPL requests exemption from the provisions of Section III G.2.d of Appendix R to 10CFR50 that requires 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.
Specifically, exemption is requested for intervening combustibles inside the primary containment for each of Turkey Point Units 3 and 4 (Fire Areas Q and P respectively).
Justification for Exemption The entire containment [sic, recte Containment] is classified as one fire area for each of Turkey Point Units 3 and 4. Each containment [sic, recte Containment] has an inside diameter of 116-feet and a free volume in excess of 1.5 million cubic feet. A primary shield wall encloses the reactor and forms the refueling cavity. There is also a secondary shield wall; a biological shield wall up to 3-feet in thickness, which encloses the primary loop.
0 Fire Safety Analysis Data Manager (4.1) 0 Turkey Point                                                      Run: 06/27/2012 14:41 0Page: 17 of 32

0                                                                                             0                                                                                                0 Attachment K - Existing Licensing Action Transition Licensing Action                 LA-10-19870812, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (IllI.G.2.d Criteria)
Evaluation By the nature of the arrangement of structures and equipment inside containment [sic, recte Containment], it is not practical to meet the literal requirements for separation with no intervening combustibles. Instead, the routing of redundant safe-shutdown circuits, fire protection features, and the location and quantity of intervening combustibles have been evaluated and considered to provide substantial assurance that a fire inside containment [sic, recte Containment] will not prevent safe shutdown of the unit.
Raceway for redundant safe-shutdown circuits tend to radiate away from the reactor and follow the containment perimeter to their respective electrical penetration rooms. The routing is generally from opposite sectors of the containment [sic, recte Containment] such that redundant safe shutdown circuits are generally separated by much more than 20-feet. To the maximum extent possible cable [sic, recte possible, cable] is rerouted to achieve at least 20-feet clear separation.
The major combustible materials are lubricating oil and cables. Lubricating oil is contained in the Reactor Coolant Pumps, which are provided with oil collection systems; CRDM and Normal Containment Cooler Fan Motors; Emergency Containment Cooler Fan Motors; various motor-operated valves and snubbers; the Reactor Coolant Drain Tank Pump Motors; and the Containment Sump Pump Motors. The Reactor Coolant Pump Motors are located in separate cubicles and are fitted with oil collection assemblies to address Appendix R, Section 111.0 requirements. The other oil sources include relatively small quantities of oil, are sparsely located and away from most safe-shutdown cable, and are not in close proximity to piping with temperatures higher than the oil flash point. Also, since personnel access to containment [sic, recte Containment] during plant operation is strictly controlled in accordance with site administrative procedures, the potential for significant quantities of transient combustibles to accumulate is reduced.
Fire protection features for these fire areas are both active and passive in nature. Passive fire protection is provided for the redundant equipment and instrumentation and their associated cables by physical separation and 1-hour rated protection. In addition, all the cables in cable trays are either coated with a fire retardant coating, Flamemastic 71A or 77, or qualified to the requirements of IEEE 383, 1974. The fire retardant coating, although not a rated fire barrier, acts as a radiant energy shield to provide a substantial level of protection against postulated exposure fires. Furthermore, the location of the trays at elevations ranging from 19'-5" to 80', outside of the secondary shield wall and the configuration of Containment [sic, recte of the Containment] Building itself precludes the possibility of cable damage due to stratification of hot gases or other products of combustion.
In the unlikely event of a fire in these areas, there are a number of available fire protection features which would serve to ensure the safe shutdown capability of the plant. Specifically, portable fire extinguishers are located inside containment [sic, recte Containment] and in the immediate vicinity of each personnel access hatch.
Also, ionization type smoke detectors are installed in the electrical penetration areas. These smoke detectors, which alarm in the Control Room would [sic, recte Room, would] ensure early detection of a fire and rapid response of the plant fire brigade.
Considering the above, it is FPL's position that strict application of Appendix R, Section III G.2.d [sic, recte IlI.G.2.d] requirement to the fire protection of redundant safe shutdown cables inside containment [sic, recte Containment] is not necessary to achieve the underlying purpose of the rule, i.e., fire protection of safe shutdown capability. Accordingly, this exemption will not result in an undue risk to the health and safety of the public and is justified under the provisions of 10CFR50.12, Section (a)(2)(ii)."
Document ID 1987-02-11 [Attachment with NRC Question and FPL Response 3] - FPL-NRC L-87-68, Request for Additional Information - Exemption to 10 CFR 50 Appendix R Evaluation This FPL submittal letter provided the NRC with additional information needed for the NRC review.It stated:
                                "NRC QUESTIONS The following information is needed to complete our evaluation of the Appendix R exemption request made in your April 25, 1986 letter."
                                "3. Attachment 4 Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                           Run: 06/27/2012 14:41       Page: 18 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-10-19870812, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (lll.G.2.d Criteria)
Evaluation Clarify the specific instances for which exemptions are requested. Identify the safe shutdown cables and equipment affected. Quantify the fire load in term of Btu per square foot and identify and give the location of specific fire hazards. Provide actual separation distances between safe shutdown cables and equipment.
Note: Attachment numbers in this RAI refer to those of the Licensee's submitted."
FPL responded with the following:
                                "3. Attachment 4 Numerous safe shutdown related components and cables are located within the Unit 3 and Unit 4 Containment Buildings. Cables are routed in conduit and in cable trays, designated SA and SB, that usually run outside the secondary shield wall at tray elevations ranging from 187" to 80'0". These trays are not typically separated by a barrier but are vertically separated by a nominal distance of 4'0" and horizontally by 1'6". In addition, redundant circuits are separated by a horizontal distance of 20' or a radiant energy shield except in cases where exemptions were granted in the March 27, 1984, SER.
The fire load in terms of BTU per square foot is 121,000 BTU per square foot for Unit 3 Containment and 117,000 BTU per square foot for Unit 4 Containment.
While the figures appear high, they are based on a small floor area (10,600 square feet) versus a large volume (1.5 million cubic feet) and a combustible inventory which includes reactor coolant pump lubricating oil which is provided with an oil collection system as required by Appendix R Section 111.0. The only other sources of concentrated combustibles are the charcoal filters and the polar crane gear case oil. The arrangement of these combustibles and a description of other containment features are included in the Turkey Point Appendix R Report submitted in April and June, 1983."
Document ID 1987-08-12 [Safety Evaluation, Sections 4.2, 4.3 and 4.4] - NRC-FPL Letter, Exemption from Requirements of Appendix R Section III.G.2 Evaluation The SER which granted this exemption states the following:
                                "4.2 Discussion Each containment building [sic, recte Containment Building] is classified as one fire area (Fire Areas P and Q) for Turkey Point Units 3 and 4. The containment building [sic, recte Containment Building] is essentially an open area with an inside diameter of 116 feet and a free volume of 1.5 million cubic feet. There are three intermediate floor levels, a primary shield wall around the reactor, and a secondary shield wall around the primary loop.
The redundant safe shutdown cables tend to run radially away from the reactor and follow the containment perimeter to their electrical penetration rooms. Although the cables are generally separated by much more than 20 feet, there are intervening combustible materials, mostly lubricating oil and other cable.
The lubricating oil is contained in the reactor coolant pumps, control rod drive mechanism, normal containment cooler fan motors, emergency containment cooler fan motors, containment sump pump motors, and various motor-operated valves and snubbers. The reactor coolant pump motors are located in separate cubicles and are fitted with oil collection assemblies to address Appendix R, Section 111.0 requirements. The other oil sources are relatively small quantities located away from most safe shutdown cables, and are not in close proximity to piping with temperatures higher than the oil flash point. Most of the area fire load of 121,000 Btu per square foot is comprised of the oil in the reactor coolant motors.
The other major source of combustible material, the cables, are either coated with a fire retardant or qualified to the requirements of IEEE Standard 383-1974. Since access to the containment [sic, recte Containment] during plant operation is strictly limited, the probability of large amounts of transient combustibles being accumulated is low.
Fire protection features include physical separation of the redundant equipment and their associated cables and 1-hour rated fire barriers. Portable fire extinguishers are located inside containment [sic, recte Containment] and in the immediate vicinity of each personnel access hatch. Ionization-type smoke detectors which alarm in Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                        Run: 06/27/2012 14:41       Page: 19 of 32
* 0 Attachment K - Existing Licensing Action Transition Licensing Action                 LA-10-19870812, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (Ill.G.2.d Criteria)
Evaluation the control room are installed in the electrical penetration area.
4.3 Evaluation The fire protection in the above fire areas does not comply with Section III.G.2 because of intervening combustibles between redundant safe shutdown components and circuitry.
The staffs concern was that a fire could cause a loss of normal safe shutdown capability. Although there is a significant fire load due to lubricating oil in the reactor coolant pump motors, the motors have an oil collection system which minimizes the possibility of a fire. The fire potential in the cables is reduced because they are either coated with a fire retardant coating or are IEEE 383 rated. The location of the cables and equipment of the mid-elevation of the large containment building [sic, recte Containment Building] will also limit their damage from hot gases caused by stratification.
The above features reduce the amount of combustible material to a low level, and along with the large volume of the containment [sic, recte Containment] provide reasonable assurance that any fire would develop slowly and have limited heat output. Therefore, it is not probable that a single fire could jeopardize both trains of redundant safe shutdown components or circuitry.
4.4 Conclusion Based on the above evaluation, the staff concludes that the licensee's existing fire protection configuration provides an equivalent level of protection to that achieved by compliance with Section [II.G.2 of Appendix R. Therefore, the licensee's request for exemption, as described above, should be granted."
Document ID PTN-FPER-08-007 Rev. 2 [Transitioning LA Bases Review &Verification] - NFPA 805 Transition Review of Existing Licensing Actions Evaluation Attachment 2 of PTN-FPER-08-007 documents verification of configurations discussed in the above Licensing Action.
Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                        Run: 06/27/2012 14:41         Page: 20 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-1 1-19870812, Appendix R Exemption from Detection and Automatic Fire Suppression Systems in Outdoor Fire Areas 76, 77, 78, 83, 87, and 90 which are Separated by a Horizontal Distance of 20 Feet or Greater (Ill.G.2.b Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section Ill.G.2.b that requires 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 with detection and suppression. Specifically, FPL requested exemption from providing fire detection and automatic fire suppression throughout outdoor fire zones for the required redundant safe shutdown related cables having 20 feet or greater horizontal separation. This exemption was granted by the NRC in an SER dated 8/12/87.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-10-010 Rev. 3 -Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                        Run: 06/27/2012 14:41      Page: 21 of 32 0                                                                                          0 40

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-1 2-19870812, Appendix R Exemption from Providing Detection and Automatic Fire Suppression in Outdoor Fire Zones 76, 77, 78, 83, 87, and 90 which are Enclosed in a Fire Barrier Having a 1-Hour Rating (III.G.2.c Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section III.G.2.c that requires the enclosure of cable and equipment and associated non-safety circuits of one redundant train in a fire barrier having a one-hour rating with detection and suppression. Specifically, FPL requested exemption from detection and automatic fire suppression throughout outdoor fire zones for the required redundant safe shutdown related cables enclosed in a fire barrier having a one-hour rating where separation of 20 feet or more cannot be provided. This exemption was granted by the NRC in an SER dated 8/12/87.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                        Run: 06/27/2012 14:41      Page: 22 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-1 3-19870812, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area AAA from Fire Area A (IllI.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request per the 04/25/86 FPL submittal letter, as supplemented by the 2/11187 letter to the NRC, provides justification for the exemption from the provisions of Section IlI.G.2.a of Appendix R to 10CFR50 that requires separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Specifically, FPL requested exemption from providing 3-hour rated penetration seals in the fire barrier separating the Gas Decay Tank Room (Fire Area AAA - Fire Zone 24) from the Auxiliary Building Corridor (Fire Area A - Fire Zone 4), and the Chemical Drain Tank Room, Laundry and Hot Shower Tank Room (Fire Area A - Fire Zone 5). This exemption was granted by the NRC in an SER dated 8/12/87.
This exemption is no longer required because according to FPL engineering evaluation PTN-FPER-1 1-010, the unsealed penetration is adequate for the hazard.
References                      Document ID PTN-FPER-1 1-010 Rev. 0 - Acceptability of Unsealed Penetrations Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                         Run: 06/27/2012 14:41      Page: 23 of 32 0                                                                                            0                                                                                            0

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-14-19870812, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area F from Fire Area A (lll.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request per the 04/25/86 FPL submittal letter, as supplemented by the 2/11/1987 letter to the NRC, provides justification for the exemption from the provisions of Section IIl.G.2.a of Appendix R to 10CFR50 that requires separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Specifically, FPL requested exemption from providing 3-hour rated penetration seals in the fire barrier separating the Purification and Radwaste Demineralizer Rooms (Fire Area F - Fire Zones 48, 49 and 50) from the Radioactive Pipeway (Fire Area A - Fire Zone 10). This exemption was granted by the NRC in an SER dated 8/12/87.
This exemption is no longer required because according to FPL engineering evaluation PTN-FPER-11-010, the unsealed penetration is adequate for the hazard.
References                       Document ID PTN-FPER-11-010 Rev. 0 - Acceptability of Unsealed Penetrations Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                         Run: 06/27/2012 14:41       Page: 24 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-15-19980224, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier Until a Horizontal Distance of 10 Feet is Attained in Roof Top Locations Excluding the Turbine Building (lll.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIl.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. Specifically, FPL requested the use of these exemptions for the raceway barriers for these outdoor applications in lieu of the 1-hour fire barriers that were previously approved in safety evaluation dated 08/12/87 (Fire Zone 118 only). This new exemption request superseded all other requests approved in the earlier 1987 Safety Evaluation. FPL specifically requested (Item 3) in the reference dated 12/12/96 an exemption for Outdoor Fire Zones permitting the use of the following (listed by exemption request number) in lieu of subsection IIl.G.2.a requirements:
3 ) Separation of cables and equipment and associated non-safety circuits of redundant trains by a 25-minute rated fire barrier until a horizontal distance of 10 feet is attained in Roof Top locations. No suppression or detection is provided.
This exemption was partially granted for outdoor fire zones 118 and 143, by the NRC in an SER dated 2/24/98.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                         Run: 06/27/2012 14:41        Page: 25 of 32 0                                                                                           0                                                                                               0

0 Attachment K - Existing Licensing Action Transition Licensing Action                 LA-16-19980224, Appendix R Exemption from ilII.G.2.a for 3 Hr Electrical Raceway Fire Barriers in the Outdoor Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120 Excluding the Turbine Building (llI.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request per the 12/12/96 FPL submittal letter provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIl.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. FPL requested the use of these exemptions for the raceway barriers for these outdoor applications in lieu of the 1-hour fire barriers that were previously approved in safety evaluations dated 3/27/84 and 08/12/87. This new exemption request superseded all other requests approved in the earlier 1984 and 1987 Safety Evaluations. This exemption was partially approved for outdoor Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120. This exemption was granted by the NRC in an SER dated 02/24/98.
In the exemption request dated 12/12/96, FPL specifically requested an exemption for Outdoor Fire Zones permitting the use of the following (listed by exemption request number) in lieu of subsection IIl.G.2.a requirements:
: 2) Separation of cables and equipment and associated non-safety circuits of redundant trains by a 25-minute rated fire barrier until a horizontal distance of 20 feet is attained. No suppression or detection is provided. This exemption was for Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120.
: 5) The existing separation of approximately 12 feet on center for the Component Cooling Water (CCW) pumps combined with fire detection and a dual-header partial-coverage suppression system for the pumps. This exemption was for Fire Zones 47 and 54.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                          Turkey Point                                                       Run: 06/27/2012 14:41         Page: 26 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-17-19981008, App R Exemption from IlI.G.2.a by a 25-Min Rated Fire Barrier until a Horizontal Distance of 20 Ft is Attained in the Outdoor Fire Zones 79-partial, 84-partial, 86, 88-partial, &89-partial Excluding the Turbine Bldg (lll.G.2.a Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request per the 12/12/96 FPL submittal letter provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section ill.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. FPL specifically requested the use of these exemptions for the raceway barriers for these outdoor applications in lieu of the 1-hour fire barriers that were previously approved in safety evaluation dated 08/12/87. This new exemption request superseded all other requests approved in the earlier 1987 Safety Evaluation. This exemption request was partially granted for Fire Zones 79-partial, 84-partial, 86, 88-partial, and 89-partial by the NRC in an SER dated 10/08/98.
In the exemption request dated 12/12/96, FPL specifically requested an exemption for Outdoor Fire Zones permitting the use of the following (listed by exemption request number) in lieu of subsection IIl.G.2.a requirements:
: 2) Separation of cables and equipment and associated non-safety circuits of redundant trains by a 25-minute rated fire barrier until a horizontal distance of 20 feet is attained. No suppression or detection is provided. This exemption was for Fire Zones 79, 84, 86 North of column line 22 and East of the A-line, 88, and 89 East of the Jc-line.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-1 0-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-10-010 Rev. 3 -Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                       Run: 06/27/2012 14:41        Page: 27 of 32 0                                                                                            0                                                                                             0

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-18-19981008, Appendix R Exemption from III.G.2.a West of the Open Turbine Building Structure Column Line A by a 1-Hour Rated Fire Barrier Until a Horizontal Distance of 20 Feet is Attained in Fire Zones 81 & 86 (lll.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request per the 12/12/96 FPL submittal letter provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section IIIlG.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. FPL specifically requested the use of these exemptions for the raceway barriers for these outdoor applications in lieu of the 1-hour fire barriers that were previously approved in safety evaluation dated 08/12/87. This new exemption request superseded all other requests approved in the earlier 1987 Safety Evaluation. This exemption request was partially granted for Fire Zones 81 and 86 by the NRC in an SER dated 10/08/98.
In the exemption request dated 12/12/96, FPL specifically requested an exemption for Outdoor Fire Zones permitting the use of the following (listed by exemption request number) in lieu of subsection Ill.G.2.a requirements:
: 1) Separation of cables and equipment and associated non-safety circuits of redundant trains west of the Open Turbine Building Structure column line A by a 1-hour rated fire barrier until a horizontal distance of 20 feet is attained. Water suppression systems are provided for the major combustible sources, however no suppression or detection is provided for the raceways. This exemption was for Fire Zones 81 and 86 West of the A-line.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-1 0-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-10-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                               Turkey Point                                                     Run: 06/27/2012 14:41       Page: 28 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-19-19990504, Appendix R Exemption from Technical Requirements for Outdoor Fire Zone 106R to the Extent that it had not Protected a Redundant Train with a 3-Hour Rated Barrier (lll.G.2.a Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section ]ll.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. Specifically, FPL requested an exemption to allow the use of separation of cables and equipment and associated non-safety circuits of redundant trains by a 25
                                -minute rated fire barrier until a horizontal distance of 10'is attained in Fire Zone 106R. This exemption had been originally denied by the NRC in the SER dated 2/24/98. This exemption was granted by the NRC in an SER dated 5/4/1999.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-10-010 Rev. 3 -Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1) 0 Turkey Point is Run: 06/27/2012 14:41 isPage: 29 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                 LA-20-19990505, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using a Minimum 1-Hour Barrier between Column Lines A and E-1(III.G.2.a Criteria)
Required Post-Transition         No Licensing Basis                  Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section III.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. Specifically, FPL requested exemption for Outdoor Fire Zones for the Open Turbine Building permitting the use of the following in lieu of subsection IIl.G.2.a requirements:
Separation of cables and equipment and associated non-safety circuits of redundant trains within the Open Turbine Building (FZs 79(partial), 80(partial), 82, 84 (partial), 85(partial), 88(partial), 91, 92, and 105) between column lines A and E-1 by a fire barrier having a 1-hour rating with automatic fixed water spray fire suppression systems, however, no fire detection is provided for this area. This exemption was originally granted by the NRC in an SER dated 12/22/98 which was then superseded by the 5/5/99 NRC SER.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-1 0-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                       Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                            Turkey Point                                                         Run: 06/27/2012 14:41        Page: 30 of 32

Attachment K - Existing Licensing Action Transition Licensing Action                LA-21-19990505, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using Separation by a 25-Minute Fire Rated Barrier between Column Lines E-1 and Jc and the Turbine Deck (lll.G.2.a Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements, of 10 CFR Part 50 Appendix R Section IIl.G.2.a that requires that cable and equipment and associated non-safety circuits of a redundant train of certain shutdown apparatus in the same fire area be separated by a fire barrier having a 3-hour rating. Specifically, FPL requested exemption for Outdoor Fire Zones for the Open Turbine Building permitting the use of the following in lieu of subsection III.G.2.a requirements:
Separation of cables and equipment and associated non-safety circuits of redundant trains within the Open Turbine Building and adjoining areas, FZs 79(partial), 84 (partial), 88(partial), 89(partial), between column lines E-1 and Jc and above the turbine operating deck, FZ 117, by a fire barrier having a 25-minute rating. Automatic wet pipe sprinkler coverage is provided between column lines E-1 and J and the Turbine Deck. The exemption was originally granted by the NRC in an SER dated 12/22/98 which was superseded by the 5/5/99 NRC SER.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                           Turkey Point                                                        Run: 06/27/2012 14:41      Page: 31 of 32 0                                                                                             0

Attachment K - Existing Licensing Action Transition Licensing Action                LA-22-20060927, Appendix R Exemption for Lack of Suppression in the Mechanical Equipment Room (Fire Zone 097) and for Lack of Suppression and Detection in the Control Room Roof (Fire Zone 106R) (III.G.3 Criteria)
Required Post-Transition        No Licensing Basis                 Exemption request provides justification for the exemption from the requirements of 10 CFR Part 50 Appendix R Section III.G.3 that requires installation of a fixed fire suppression system in the area, for which an alternate shutdown capability independent of cables, systems or components in that area, is provided. Specifically, FPL requested exemption from providing a fixed fire suppression system in the Mechanical Equipment Room and fixed fire suppression and detection systems in the Control Room Roof (both fire zones are treated as part of Fire Area MM, the Main Control Room). This exemption was granted by the NRC in an SER dated 9/27/06.
This exemption is no longer required because the fire risk evaluation, PTN-BFJR-10-010, has found that the fire area is compliant with NFPA 805 Section 4.2.4.
References                      Document ID PTN-BFJR-1 0-010 Rev. 3 - Turkey Point Nuclear Fire Risk Evaluations Fire Safety Analysis Data Manager (4.1)                                                        Turkey Point                                                       Run: 06/27/2012 14:41        Page: 32 of 32

Florida Power & Light       Attachment L - NFPA 805 Chapter 3 Requirements for Approval L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
20 Pages Attached Page L-1 Revision 00                                                                    Page L-1

Florida Power & Light             Aftachment L - NFPA 805 Chapter 3 Requirements for Approval 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:
(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).
FPL requests formal approval of performance-based exceptions to requirements in Chapter 3 of NFPA 805 as follows:
Page L-2 Revision 00                                                                            Page L-2

Florida Power & Light             Affachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request I NFPA 805 Section 3.3.3 NFPA 805 Section 3.3.3 states:
    "InteriorFinishes. Interior wall or ceiling finish classification shall be in accordance with NFPA 101I, Life Safety Code, requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class / interior floor finishes."
Turkey Point Plant (PTN) utilizes an epoxy floor coating system that does not meet the exact requirements of NFPA 805 Section 3.3.3.
NFPA 101 requirements for interior floor finishes state that the floor finish shall be characterized by a critical radiant flux not less than 0.45 W / cm 2 . In addition, the NRC issued Information Notice (IN) 2007-26 to address the combustibility of epoxy floor coatings at commercial nuclear power plants. Per IN 2007-26, the NRC defined a non-combustible material as:
     " material having a structural base of noncombustible material,
     " an ASTM E-84 flame spread rating not higher than 50,
     " and is not over 1/8" (125 mils) thick NFPA 805 has re-defined the IN 2007-26 definition of non-combustible material to limited combustible: "Materialthat, in the form in which it is used, has a potential heat value not exceeding 3500 Btu/lb (8141 kJ/kg) and either has a structuralbase of noncombustible material with a surfacing not exceeding a thickness of 1/8 in. (3.2 mm) that has a flame spread rating not greaterthan 50, or has anothermaterial having neithera flame spread rating greaterthan 25 nor evidence of continued progressive combustion, even on surfaces exposed by cutting through the materialon any plane."
NFPA 805 defines non-combustible material as "Materialthat, in the form in which it is used and under the conditions anticipated,will not ignite, bum, support combustion, or release flammable vapors when subjected to fire or heat."
In some areas, Turkey Point Plant (PTN) utilizes an epoxy floor coating system. The critical radiant flux of the epoxy floor coating system is not readily available, but a previous licensee evaluation of the combustibility of the epoxy floor coatings was performed. The PTN evaluation was to determine if the epoxy floor coatings used had a significant combustible value to be included in the fire protection program. The PTN floor coating specifications were reviewed and it was determined that the maximum concrete floor coating thickness used was 62 mils. Review of manufacturer data of the coatings used indicated that the ASTM E-84 flame spread value for a 49 mils thick epoxy floor coating system similar to the PTN coating system is 30.
The basis for the approval request of this deviation is:
     " The form in which the epoxy floor coating is used and conditions anticipated would meet the definition of a limited combustible material
     " The maximum floor coating thickness used at PTN is 62 mils (significantly less than 1/8" or 125 mils).
Reviion                                                    Pag L-Revision 0                                                                             Page L-3

Florida Power & Light              Attachment L - NFPA 805 Chapter 3 Requirements for Approval
    "    The ASTM E-84 flame spread value of the PTN epoxy coating system at the slightly larger thickness is anticipated to have a flame spread rating less than 50.
    " The epoxy floor coating is at the floor level. The ASTM E-84 test is conducted with the material on the ceiling of a tunnel. This configuration would allow the flame to directly impinge on the ceiling surface, enhancing flame spread. With the material on the floor, the heat flux to the surface is much less than would be expected in the ceiling configuration since the convective flame is directing the heat away from the surface. This would mean that the overall flame spread would be expected to be much less, even with a slightly greater thickness.
    " The epoxy coating with the aforementioned criteria would not result in propagation across barriers or between redundant success paths.
Nuclear Safety and Radiological Release Performance Criteria:
The use of epoxy floor coating does not affect nuclear safety as it meets the definition of a limited combustible material. Application of epoxy floor coatings is controlled via PTN specifications to ensure that the amount of material does not add appreciable amounts of combustible material to the plant. The epoxy coating would not result in propagation across barriers or between redundant success paths. Therefore there is no impact on the nuclear safety performance criteria.
The use of epoxy floor coatings has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the floor coating materials. The floor coatings do not change the results of the radiological release evaluation performed that concluded that potentially contaminated water is contained and smoke is monitored. Floor coatings do not add additional radiological materials to the area or challenge systems boundaries.
Safety Margin and Defense-in-Depth:
The epoxy floor coatings meet the definition of a limited combustible material.
Therefore, the safety margin inherent in the analysis for the fire event has been preserved. The use of epoxy floor coating does not affect safety margin as it meets the definition of a limited combustible material. The floor coating materials were evaluated to have a negligible effect on combustibility. Application of epoxy floor coatings is controlled via a PTN procedure. The areas with epoxy floor coatings have been analyzed in their current configuration. The precautions and limitations on the use of these materials do not impact the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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). The use of epoxy floor coatings do not affect echelons 1, 2, and 3. The use of epoxy floor coatings does not directly result in compromising Revision 0                                                                             Page L-4

Florida Power & Light              Attachment L - NFPA 805 Chapter 3 Requirements for Approval automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for approval of the use of epoxy floor coatings as compliant with NFPA 805 Section 3.3.3.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-5 Revision 00                                                                             Page L-5

Florida Power & Light              Attachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 2 NFPA 805 Section 3.3.5.1 NFPA 805 Section 3.3.5.1 states:
     "Wiring above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be listed for plenum use, routed in armoredcable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers."
PTN has identified wiring above suspended ceilings that may not comply with the requirements of this code section. This wiring may be non-listed video/communication/data cables.
Historically, Appendix A to BTP 9.5-1 Guidance Section D.1 (f) has the requirement "Suspendedceilings and their supports should be of noncombustible construction.
Concealed spaces should be devoid of combustibles." PTN's response was that "Suspendedceilings and their supports are non-combustible to the extent possible. No significantcombustibles are located in concealed spaces."
The areas at PTN currently with suspended ceilings inside the NFPA 805 defined power block include:
     " Auxiliary Building office areas (Counting Room, Hot Lab)
     " Control Room/associated offices These areas are not risk significant with the exception of the Control Room. Current PTN cable construction for power and control cables are IEEE-383 (or equivalent) or routed in metal conduit. Original power and control cables may not have been IEEE-383 or equivalent. Where non IEEE-383 cables are used in cable trays, the use of fire retardant coatings was applied in certain areas. Field walkdowns and review of design drawings confirmed that there are no power/control cables above the ceiling in the Auxiliary Building Office areas but there are cable trays with power and control cables above the Control Room ceiling. The power cables are 120VAC or 125VDC. The cable trays are either sprayed with a fire retardant coating (Flamemastic) or are new trays with IEEE-383 cables. NFPA 805 FAQ 06-0022 identified acceptable electrical cable construction tests. Plenum rated cable is tested to NFPA 262. The FAQ concluded that the NFPA 262 test is equivalent to the IEEE-383 test. Therefore, IEEE cable is inherently equivalent to plenum rated cable and acceptable to be routed above suspended ceilings. In addition, those cables sprayed with a flame retardant coating do not pose a risk as the combustible element of the cable jacketing has been nullified by the coating.
Therefore the cables in question are video/communication/data cables which have been field routed above suspended ceilings. These cables may not be plenum rated.
Video/communication/data cables are low voltage. These low voltage cables are not generally susceptible to shorts which would result in a fire.
The basis for the approval request of this deviation is:
* The wiring above ceilings in offices, laboratories, etc. do not pose a hazard:
o Low voltage is not susceptible to shorts causing a fire.
Revision 0                                                                             Page L-6

Florida Power & Light             Attachment L - NFPA 805 Chapter 3 Requirements for Approval o Power and control cables are protected (plenum rated equivalent) per this code section or sprayed with a flame retardant coating.
o By eliminating cables with the potential shorts, this eliminates ignition sources and therefore the jacketing of cable is not relevant.
Nuclear Safety and Radiological Release Performance Criteria:
The location of wiring above suspended ceilings does not affect nuclear safety. Power and control cables comply, or comply with the intent of this section. Other wiring, while it may not be in armored cable, in metallic conduit, coated in a flame retardant material, or plenum rated, is low voltage cable not susceptible to shorts that would result in a fire.
Therefore there is no impact on the nuclear safety performance criteria.
The location of cables above suspended ceilings has no impact on the radiological release performance criteria. The radiological review was performed based on the potential location of radiological concerns and is not dependent on the type of cables or locations of suspended ceilings. The cables do not change the results of the radiological release evaluation performed that concluded that potentially contaminated water is contained and smoke is monitored. The cables do not add additional radiological materials to the area or challenge systems boundaries.
Safety Margin and Defense-in-Depth:
Power and control cables meet the requirements or the intent of this requirement. Other wiring, while it may not be in armored cable, in metallic conduit, coated in a flame retardant material, or plenum rated, is low voltage cable not susceptible to shorts that would result in a fire. These areas with video/communication/data cables have been analyzed in their current configuration. Therefore, the inherent safety margin and conservatisms in these analysis remain unchanged.
The three echelons of defense-in-depth 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). The prior introduction of non-listed video/communication/data cables routed above suspended ceilings does not impact fire protection defense-in-depth. Echelon 1 is maintained by the cable installation procedures documenting the requirements of NFPA 805 Section 3.3.5.1. The introduction of cables above suspended ceilings does not affect echelons 2 and 3. The video/communication/data cables routed above suspended ceilings does not directly result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the use of non-listed video/communication/data cables routed above suspended ceilings.
Page L-7 Revision 00                                                                          Page L-7

Florida Power & Light              Aftachment L - NFPA 805 Chapter 3 Requirements for Approval The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-8 Revision 00                                                                             Page L-8

Florida Power & Liqht               Affachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 3 NFPA 805 Section 3.3.5.2 NFPA 805 Section 3.3.5.2 states:
     "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."
The use of PVC piping for underground embedded conduit is permitted by a PTN specification for electrical raceway installations. Polyvinyl Chloride (PVC) or High Density Polyethylene (HDPE) type ducts (conduits) are permitted when embedded in compacted sand or reinforced concrete. In addition, some PVC conduit was found in reinforced concrete wall. The PVC/HDPE conduit is embedded within a non-combustible enclosure which provides protection from mechanical damage and from damage resulting from either an exposure fire or from a fire within the conduit impacting other targets.
The basis for the approval request of this deviation is:
     " The PVC/HDPE conduit, while a combustible material, is not subject to flame/heat impingement from an external source which would result in structural failure, contribution to fire load, and damage to the circuits contained within where the conduit is embedded in concrete or compacted sand.
     " Failure of circuits within the conduit resulting in a fire would not result in damage to external targets.
Nuclear Safety and Radiological Release Performance Criteria:
The use of PVC/HDPE conduit in embedded locations does not affect nuclear safety as the material in which conduits are run within an embedded location is not subject to the failure mechanisms potentially resultant in circuit damage or resultant damage to external targets. Therefore there is no impact on the nuclear safety performance criteria.
The use of PVC/HDPE conduits in embedded installations has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the type of conduit material. The conduit material does not change the radiological release evaluation performed that concluded that potentially contaminated water is contained and smoke is monitored. The conduits do not add additional radiological materials to the area or challenge systems boundaries that contain such as the PVC/HDPE conduits are embedded.
Safety Margin and Defense-in-Depth:
The PVC/HDPE conduit material is embedded in a non-combustible configuration. The material is protected when embedded from mechanical damage and from damage resulting from either an exposure fire or from a fire within the conduit impacting other Revision 0                                                                              Page L-9

Florida Power& Light               Aftachment L - NFPA 805 Chapter 3 Requirements for Approval targets. The areas with PVC/HDPE conduit have been analyzed in their current configuration. The precautions and limitations on the use of these materials do not impact the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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). The use of PVC/HDPE conduits in embedded installations does not impact fire protection defense-in-depth. The PVC/HDPE conduit in embedded installations does not affect echelons 1, 2, and 3. The PVC/HDPE conduits do not directly result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the use of PVC/HDPE conduits in embedded installation.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-1O Revision 00 Revision                                                                             Page L-10

Florida Power & Light             Affachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 4 NFPA 805 Section 3.3.12(1)
NFPA 805 Section 3.3.12(1) states:
    "The oil collection system for each reactorcoolant pump shall be capable of collecting lubricatingoil from all potentialpressurized and nonpressurizedleakage sites in each reactorcoolant pump oil system."
The PTN oil collection system is designed and was reviewed in accordance with 10 CFR 50, Appendix R, Section 111.0 to collect leakage from credible pressurized and nonpressurized leakage sites in the reactor coolant pump oil system. This may not include collection of oil mist as the result of pump/motor operation. Oil misting is not leakage due to equipment failure, but inherent in the operation of large open motors. It is normal for large motors to lose some oil through seals and the oil to potentially become 'atomized' by ventilation air flow. This atomized oil mist can then collect on surfaces in the vicinity of the reactor coolant pump as the pump design is not completely sealed to permit airflow for cooling. The oil mist resulting from normal operation will not adversely impact the ability of a plant to achieve and maintain safe shutdown even if ignition occurred. There are redundant reactor coolant pumps and they are not required to achieve and maintain safe shutdown.
In addition Generic Letter 86-10, Response to Industry Questions, dated April 24, 1986; Question 6.2 (shown below) discussed oil dripping. The response concluded that there was no concern with oil consumption (which is an oil misting phenomenon) but the concern was with an oil fire started from a pressurized leakage point and/or spilled leakage.
    "QUESTION 6.2 It would appearthat a literal reading of Section III. 0 regardingthe oil collection system for the reactorcoolant pump could be met by a combination of seismically designed splash shields and a sump with sufficient capacity to contain the entire lube oil system inventory. If the reactorcoolant pump is seismically designed and the nearby piping hot surfaces are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, does this design concept conform to the requirements of the rule?

===RESPONSE===
If the reactorcoolantpump, including the oil system, is seismically designed and the nearby hot surfaces of piping are protected by seismically designed splash shields such that any spilled lube oil would contact only cold surfaces, and it could be demonstratedby engineering analysis that sump and splash shields would be capable of preventing a fire during normal and design basis accident conditions, the safety objective of Section II.0 would be achieved. Such a design concept would have to be evaluated under the exemption process. The justification for the exemption should provide reasonableassurance that oil from all potential pressurizedand unpressurized leakage points would be safely collected and drained to the sump. The sump should be shown capable of safely containing all of Revision 0                                                                           Page L-11

Florida Power & Light              Aftachment L - NFPA 805 Chapter 3 Requirements for Approval the anticipatedoil leakage. The analysis should verify that there are no electric sources of ignition."
The PTN response to 10 CFR Part 50, Appendix R Requirements stated, "The system collects oil from the lift pump and piping, lube oil coolerand associated piping drain, upperand lower oil reservoirswitch assemblies, oil fill line flexible hose connections, flanged connections on oil lines and the lube oil reservoirs."
Historically, there have been no reported fires attributed to oil misting based on normal operation in the industry. Fires have occurred due to oil leakage from equipment failure such as cracked welds on piping or inadequate collection pan design or maintenance.
PTN does not have a history of significant oil loss from the reactor coolant pumps as a result of oil misting or oil leakage that is not contained by the properly designed and installed oil leakage collection system.
The basis for the approval request of this deviation is:
    " The oil collection system is designed to collect leakage from credible pressurized and nonpressurized leakage sites in the reactor coolant pump oil system.
    " Oil misted from normal operation is not leakage; it is normal motor oil consumption.
* Oil misted from normal operation does not significantly reduce the oil inventory.
The oil historically released as misting does not account for an appreciable heat release rate or accumulation near potential ignition sources or non-insulated reactor coolant piping.
    " The reactor coolant pumps use an oil of a high flash point, over 400 degrees Fahrenheit.
    " There are redundant reactor coolant pumps and they are not essential to achieve or maintain fire safe shutdown.
Nuclear Safety and Radiological Release Performance Criteria:
The oil mist resultant from normal operations will not adversely impact nuclear safety.
There are redundant reactor coolant pumps and they are not required to achieve or maintain post-fire safe shutdown. Therefore there is no impact on the nuclear safety performance criteria.
The potential for oil mist from the reactor coolant pumps has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials. The entire Reactor Building in which the reactor coolant pumps are located is an environmentally sealed radiological area. The oil mist does not add additional radiological materials to the area or challenge system boundaries.
Safety Margin and Defense-in-Depth:
The oil mist resultant from normal operation will not adversely impact the ability of a plant to achieve and maintain post-fire safe shutdown even if ignition occurred. There are redundant reactor coolant pumps, however the reactor coolant pumps are not Revision 0                                                                            Page L-12

Florida Power & Light              Affachment L - NFPA 805 Chapter 3 Requirements for Approval required to achieve and maintain fire post-fire safe shutdown. The reactor buildings have been analyzed in their current configuration. The precautions and limitations on potential oil misting does not impact the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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). The potential for oil mist from reactor coolant pumps does not impact fire protection defense-in-depth. Echelon 1 is maintained by the oil collection system and reactor coolant pump design. The introduction of small amounts of oil misting does not affect echelons 2 and 3. The potential for oil mist from the reactor coolant pumps does not result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the potential of oil misting from the reactor coolant pumps due to normal motor consumption not captured by the oil collection system designed for pressurized and non-pressurized leakage and spillage.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-13 Revision 00                                                                           Page L-13

Florida Power & Light             Affachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 5 NFPA 805 Sections 3.5.5 and 3.9.4 NFPA 805 Section 3.5.5 states:
    "Eachpump and its driver and controls shall be separatedfrom the remaining fire pumps and from the rest of the plant by rated fire barriers."
NFPA 805 Section 3.9.4 states:
    "Diesel-drivenfire pumps shall be protectedby automatic sprinklers."
There are no rated fire barriers separating the electric and the diesel-driven fire pumps or respective controllers at PTN. In addition, the diesel-driven fire pump (DDFP) is not provided with automatic sprinkler protection.
The electric fire pump, diesel driven fire pump, and associated controllers are located in the northeast corner of the protected area near the Raw Water Tanks and water treatment area. The diesel driven fire pump is located in an enclosed structure for weather protection while the electric fire pump is located outside. The electric fire pump is located approximately 50 feet to the west of the diesel driven fire pump. The fire pump controllers are located outside adjacent to the respective fire pumps. The controllers are separated by more than 20'. This area of the yard has the Raw Water Tanks, the jockey pumps between the two fire pumps and miscellaneous other service water pumps to the west of the electric fire pump. There are no significant intervening combustibles between the pumps/controllers. These areas are a significant distance (greater than 100 feet) from the next nearest power block structure, the Intake Structure.
Each fire pump is individually capable of providing the required capacity of the fire protection water system. The diesel driven fire pump has a nominal 500 gallon fuel tank located outside but adjacent to the diesel driven fire pump house within a containment area in the event of a spill or rupture. This tank provides over eight hours of operational time. The electric fire pump is powered via the plant vital power system and can be powered by the emergency diesel generator system.
The basis for the approval request of this deviation is:
    " The diesel driven and electric fire pumps are spatially separated
    " There are no credible fire scenarios that would impact both fire pumps
    " Loss of one fire pump would not impact the ability to provide 100% of the required fire water demand. In addition, there are contingency plans for alternative fire water supply capability.
    " Reliable backup power supply is available to the electric fire pump in the event of a loss of offsite power and the diesel driven fire pump is impaired.
Nuclear Safety and Radiological Release Performance Criteria:
The lack of fire rated separation between fire pumps and lack of automatic suppression over the diesel-driven fire pump does not affect nuclear safety as the fire pumps are not relied upon for nuclear safety functions. Each fire pump individually has the ability to Revision 0                                                                          Page L-14

Florida Power & Light              Affachment L - NFPA 806 Chapter 3 Requirements for Approval supply the required fire water and is not relied upon for other water requirements.
Therefore there is no impact on the nuclear safety performance criteria.
The lack of fire rated separation between fire pumps and lack of automatic suppression over the diesel-driven fire pump has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the location of the fire pumps. The location of the fire pumps does not change the radiological release evaluation performed that concluded that potentially contaminated water is contained and smoke is monitored. The configuration of the fire pumps do not add additional radiological materials to the area or challenge systems boundaries.
Safety Margin and Defense-in-Depth:
The lack of fire rated separation between fire pumps and lack of automatic suppression over the diesel-driven fire pump does not negate the ability to supply the required fire water in a fire event. The nuclear safety analysis does not credit both of the fire pumps simultaneously, nor required suppression over the diesel fire pump. The use of these systems has been defined by the limitations of the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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 3 is maintained by adequate separation of the fire pumps to ensure one fire pump is operable if a fire affects another pump. The location of the fire pumps does not affect echelons 1 and 2. The lack of fire rated separation between fire pumps and lack of automatic suppression over the diesel-driven fire pump does not result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the lack of fire rated separation between fire pumps and lack of automatic suppression over the diesel-driven fire pump.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-15 Revision 0 Revision   0                                                                          Page L-15

Florida Power & Liaht              Aftachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 6 NFPA 805 Sections 3.8.1(2) and 3.9.3 NFPA 805 Section 3.8.1 states:
     "Alarm initiatingdevices shall be installedin accordance with NFPA 72, National Fire Alarm Code. Alarm annunciationshall allow the proprietaryalarm system to transmit fire-relatedalarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnelassigned to the proprietaryalarm station shall be permitted to have other duties. The following fire-relatedsignals shall be transmitted."
NFPA 805 Section 3.8.1(2) states:
     "Actuation of any fixed fire suppressionsystem" NFPA 805 Section 3.9.3 states:
     "All alarms from fire suppression systems shall annunciatein the control room or othersuitable constantly attended location."
PTN has some automatic water-based suppression systems in the nuclear power block with local alarm only. The following suppression systems are not provided with an alarm that annunciates in a constantly attended location:
     " Wet pipe sprinkler system for Unit 3 Turbine Building
     " Wet pipe sprinkler system for Unit 4 Turbine Building The lack of an alarm in a constantly attended location does not prevent the suppression system from actuating and performing the design function to control a fire. Actuation of a water-based suppression system would result in the fire pump start. The fire pump start alarms in the Control Room which is a constantly attended location. Start of the fire pump without direct alarm indication would initiate an investigation as to the cause of the start. In addition, each suppression system listed above has a local alarm.
These suppression systems are located in the open areas of the Turbine Building where personnel nearby would hear the alarm and report it per station procedures.
The basis for the approval request of this deviation is:
     " The suppression system will actuate and perform the designed function to control the fire
     " A fire pump would start on low pressure in the fire main resulting in an alarm to the Control Room
     " Either alarm via the fire pump or report via personnel hearing the local alarm would result in fire brigade response Nuclear Safety and Radiological Release Performance Criteria:
The lack of automatic water-based suppression system alarms to a constantly attended location does not affect nuclear safety as the suppression systems will actuate and perform the designed function to control the fire. Therefore there is no impact on the nuclear safety performance criteria.
Revision 0                                                                           Page L-16

Florida Power & Light              Affachment L - NFPA 805 Chapter 3 Requirements for Approval The lack of automatic water-based suppression system alarms to a constantly attended location has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the type of alarms on the suppression system. The type of alarms on the suppression systems do not change the radiological release evaluation results that concluded that potentially contaminated water is contained and smoke is monitored.
The type of alarms on the suppression systems do not add additional radiological materials to the area or challenge systems boundaries.
Safety Margin and Defense-in-Depth:
The lack of automatic water-based suppression system alarms to a constantly attended location does not negate the ability of the suppression system to actuate and perform the intended design function in a fire event. The use of these systems has been defined by the limitations of the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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). The lack of automatic water-based suppression system alarms to a constantly attended location does not affect echelons 1, 2, and 3. The lack of alarm to a constantly attended location does not compromise automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the lack of alarms on automatic water-based suppression systems to a constantly attended location.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
(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).
Page L-17 Revision 00                                                                           Page L-17

Florida Power & Light               Affachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 7 NFPA 805 Section 3.3.8 and NFPA 30 Sections 2-1.1.1, 2-2.4.2, 2-4.4.3, & 2-5.2 NFPA 805 Section 3.3.8 states:
     "Bulk storage of flammable and combustible liquids shall not be permitted inside structures containingsystems, equipment, or components importantto nuclear safety. As a minimum, storage and use shall comply with NFPA 30, Flammable and Combustible Liquids Code."
NFPA 30 (1976 Edition) Section 2-2.1.1 states:
     "Every aboveground tank for the storage of Class II or Class IliA liquids, except as provided in 2-2.1.2 and those liquids with boil-over characteristicsand unstable liquids, operating at pressuresnot in excess of 2.5 psig and designed with a weak roof-to-shell seam or equipped with emergency venting which will not permit pressures to exceed 2.5 psig, shall be located in accordance with Table 2-1." [See code for tank type definition and table]
NFPA 30 (1976 Edition) Section 2-2.4.2 states:
     "Normalvents shall be sized either in accordancewith: (1) the American Petroleum Institute StandardNo. 2000, Venting Atmospheric and Low-Pressure Storage Tanks, 1973; or (2) other accepted standard;or (3) shall be at least as large as the filling or withdrawal connection, whichever is largerbut in no case less than 1!1 inch nominal inside diameter."
NFPA 30 (1976 Edition) Section 2-4.4.3 states:
     "Tanks for storage of Class I or Class II liquids inside of buildings shall be provided with either:
(a) a normally closed remotely activated valve, (b) an automatic-closingheat-activated valve, or (c) anotherapproved device on each liquid transferconnection below the liquid level, except for connections used for emergency disposal, to provide for quick cut-off in the event of a fire in the vicinity of the tank This function can be incorporatedin the valve requiredin 2-4.4.2, and if a separate valve, shall be located adjacent to the valve required in 2-4.4.2."
NFPA 30 (1976 Edition) Section 2-5.2 states:
     "When tanks are supportedabove the foundations, tank supports shall be installed on firm foundations. Supports for tanks storing Class I, Class II, or Class IliA liquids shall be of concrete, masonry or protected steel. Single wood timber supports (not cribbing)laid horizontally may be used for outside above-ground tanks if not more than 12 inches high at theirlowest point."
The NFPA 30 code compliance evaluation identified some non-conformances. The following discussion includes those sections which PTN is requesting NRC approval of the deviation.
Page L-18 Revision 00 Revision                                                                               Page L-18

Florida Power & Light              Affachment L - NFPA 805 Chapter 3 Requirements for Approval NFPA 30 Section 2-2.1.1:
NFPA 30 states that tanks shall be a specific distance from the nearest building. In this case, the distance is five feet. The diesel driven fire pump (DDFP) fuel oil storage tank is located approximately two feet from the DDFP building. The fuel oil storage tank is located within a containment area capable of containing the entire contents of the tank.
Loss of the fuel oil storage tank would prohibit the use of the DDFP. The electric fire pump would be operational since the pumps are located approximately 50 feet apart, controllers are separated by more than 20 feet, and therefore would not be affected by the loss of the DDFP. Each pump individually is capable of providing the required water capacity. This area is a significant distance from the nearest power block structure, the Intake Structure.
The basis for the approval request of this deviation is:
     " The only structure within five feet of the nearest building is the DDFP building.
     " The fuel tank is dedicated to serve the DDFP.
     " Loss of the DDFP fuel oil storage tank would result in loss of the DDFP, therefore location of the DDFP fuel oil storage tank adjacent to the DDFP building is not relevant.
     " The electric fire pump has the capacity to provide the required fire water supply in the event of loss of the DDFP.
NFPA 30 Section 2.2.4.2:
NFPA 30 states that the'vent line should be equal to or greater than the largest filling or withdrawal connection, or sized per accepted standards. The DDFP fuel oil storage tank is equipped with a 1-1/2" vent line and a 2" fill line. The tank was purchased and installed with the DDFP but there is no record of a sizing calculation. The DDFP fuel oil storage tank has a capacity of approximately 500 gallons. The tank is typically filled using the local fill connection small transfer pump and portable tank; the 2" direct fill line is not frequently used. The rate at which the tank is filled, provided the small capacity, will not result in rupture of the tank. Filling operations require personnel be present at the tank, who stop the re-filling operation as necessary thereby preventing overflow or damage to the tank.
The basis for the approval request of this deviation is:
     " The DDFP tank is a small volume tank that is refilled in accordance with station procedures which would not result in damage to the tank.
     " The electric fire pump has the capacity to provide the required fire water supply in the event of loss of the DDFP.
NFPA 30 Section 2.4.4.3:
NFPA 30 states that for tanks inside buildings with Class II or lilA liquids shall have a valve capable of quick cut-off in the event of a fire in the vicinity of the tank. The Unit 3 Emergency Diesel Generator (EDG) day tanks do not have valves that meet the criteria of this section. Each Unit 3 EDG day tank is located in a small dedicated room in the Unit 3 EDG building, above their respective EDG. Each room is segregated by three-hour rated fire barriers, a pre-action suppression system, and containment in the event Revision 0                                                                            Page L-19

Florida Power & Light              Aftachment L - NFPA 805 Chapter 3 Requirements for Approval of a tank leak. Each day tank is equipped with two drain lines which each have a normally closed shutoff valve. The other line below the liquid level is the discharge line to the engine driven fuel pump. The discharge line to the fuel pump is equipped with a normally opened ball valve outside of the day tank room.
The basis for the approval request of this deviation is:
* The tanks are located in small individual and dedicated segregated rooms with automatic fire suppression, rated barriers, and limited combustibles NFPA 30 Section 2.5.2:
NFPA 30 states that when tanks are supported above the foundation the supports shall be of concrete, masonry or protected steel. The DDFP fuel oil storage tank is supported by unprotected steel legs. The fuel oil storage tank is located within a containment area capable of containing the entire contents of the tank. Fireproofing on the supports could prevent an exposure fire from causing structural failure of the supports and the fuel from the tank adding to the fire. The only exposure to the tank supports is from the tank itself, thereby additional protection would not be of benefit. Loss of the fuel oil storage tank would prohibit the use of the DDFP. The electric fire pump would be operational since the pumps are located approximately 50 feet apart,. controllers are separated by more than 20 feet, and each pump individually is capable of providing the required water capacity.
The basis for the approval request of this deviation is:
     " The DDFP fuel oil storage tank is located within a containment area.
     " There are no exposures to the tank other that itself, therefore additional protection would not assure survival of the tank.
     " The electric fire pump has the capacity to provide the required fire water supply in the event of loss of the DDFP.
Nuclear Safety and Radiological Release Performance Criteria:
The deviations from NFPA 30 regarding flammable liquid storage tanks do not affect nuclear safety as the DDFP is not relied upon for nuclear safety functions. The electric fire pump, in the event of loss of the DDFP, has the ability to supply the required fire water and is not relied upon for other water requirements. Therefore there is no impact on the nuclear safety performance criteria.
The deviations from NFPA 30 regarding flammable liquid storage tanks have no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the location of the fire pumps. The deviations from NFPA 30 do not change the radiological release evaluation results that concluded that potentially contaminated water is contained and smoke is monitored. The deviations do not add additional radiological materials to the area or challenge systems boundaries.
Page L-20 Revision 00 Revision                                                                            Page L-20

Florida Power & Light              Affachment L - NFPA 805 Chapter 3 Requirements for Approval Safety Margin and Defense-in-Depth:                                                             0 The deviations from NFPA 30 regarding flammable liquid storage tanks do not negate the ability to supply the required fire water in a fire event. The nuclear safety analysis does not credit both of the fire pumps simultaneously. The use of these systems has been defined by the limitations of the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.
The three echelons of defense-in-depth 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 3 is maintained by availability of the electric fire pump which has the capacity to provide the required fire water supply in the event of loss of the DDFP. The deviations from NFPA 30 do not affect echelons 1 and 2. The deviations do not result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability.

==
Conclusion:==

NRC approval is requested for the deviations to NFPA 30 regarding configurations described above.
The engineering analysis performed determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:                    0 (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).
Page L-21 Revision 0 0                                                                          Page L-21

Florida Power & Light                     Attachment M- License Condition Changes M. License Condition Changes 3 Pages Attached Page M-1 Revision 00                                                               Page M-1

Florida Power & Light                                  Attachment M - License Condition Changes Replace the current FPL Turkey Point Nuclear Plant Unit 3 and Unit 4 fire protection License Conditions 3.D with the standard license condition based upon Regulatory 0
Position 3.1 of RG 1.205.
FPL Turkey Point Nuclear Plant Unit 3 and Unit 4 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated (and supplements dated    _)                and as approved in the safety evaluation report dated                   (and supplements dated                    ). Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 X10-7 /year (yr) for CDF and less than 1xl0-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Other Changes that May Be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.
Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Reviion                                                     Pag M-Revision 0                                                                              Page M-2

Florida Power & Light                                 Attachment M - License Condition Changes Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
        " Fire Alarm and Detection Systems (Section 3.8);
        " Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
        " Gaseous Fire Suppression Systems (Section 3.10); and,
         " Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated                   to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Transition License Conditions
: 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2. The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S, Table S-2.
: 3. The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
FPL Turkey Point Nuclear Plant Unit 3 and Unit 4 Renewed Facility Operating Licenses DPR-31 and DPR-41 License Conditions 3.D state:
Page M-3 Revision 00                                                                            Page M-3

Florida Power & Light                               Affachment M - License Condition Chanqes FPL shall implement and maintain in effect all provisions of the approved Fire Protection Programas described in the Updated Final Safety Analysis Report (UFSAR) for Turkey Point Units 3 and 4 and as approved in the Safety Evaluation Report (SER) dated March 21, 1979 and supplemented by NRC letters dated April 3, 1980, July 9, 1980, December 8, 1980, January26, 1981, May 10, 1982, March 27, 1984, April 16, 1984, August 12, 1987, and by Safety Evaluations dated February 25, 1994, February24, 1998, October 8, 1998, December 22, 1998, May 4, 1999, and May 5, 1999, 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 maintain safe shutdown in the event of a fire.
It is FPL's understanding that implicit in the revocation of this license condition, all prior fire protection program SERs and commitments have been superseded in their entirety by the revised license condition.
No other license conditions need to be revised or superseded.
FPL implemented the following process for determining that these are the only license conditions required to be either revised or superseded to implement the new fire protection program which meets the requirements in 10 CFR 50.48(a) and 50.48(c):
A review was conducted of the FPL Turkey Point Nuclear Plant Unit 3 and Unit 4 Renewed Operating License Nos. DPR-31 and DPR-41. The review was performed by reading the Operating Licenses and performing electronic searches. Outstanding LARs that have been submitted to the NRC were also reviewed for potential impact on the license conditions.
Revision 0                                                                            Page M-4

Florida Power & Light                    Attachment N - Technical Specification Changes N. Technical Specification Changes 19 Pages Attached Page N-I Revision 00                                                                     Page N-1

Florida Power & Light                         Aftachment N - Technical Specification Changes Delete the following Technical Specification:
Section 6.8.1 Written procedures shall be established, implemented, and maintained covering the following activities:
: f. Fire Protection Program implementation No other Technical Specifications need to be revised or deleted.
PTN implemented the following process for determining that these are the only Technical Specifications required to be revised or deleted to implement the new fire protection program which meets the requirements in 10 CFR 50.48(a) and 50.48(c).
* A review was conducted by reading the Technical Specifications and performing electronic searches. Outstanding Technical Specification changes that have been submitted to the NRC were also reviewed for potential impact on the license conditions.
PTN determined that these changes to the Technical Specifications are adequate for adoption of the new fire protection licensing basis, for the following reasons.
* The requirement for establishing, implementing, and maintaining fire protection procedures is contained in the regulation (10 CFR 50.48(a) and 50.48(c) NFPA 805 Chapter 3).
The mark ups and retypes follow.
Page N-2 Revision 00 Revision                                                                             Page N-2

Florida Power & Light              Attachment N - Technical Specification Changes FACILITY OPERATING LICENSE, TECHNICAL SPECIFICATIONS MARKUP 7 Pages Attached Page N-3 Revision 00 Revision                                                                  Page N-3

PTN Unit 3 Facility Operating License and Technical Specification mark-up LAR Attachment N 4
D.      ".
Insert Attachment       ICft      dated April 3,1080, July , 1080, Dtcmbn iss                                             8, 1980, Janfato       26, 1081, May 10, 1082, March 27, 1084, April 16, 1084, August 12, 1987, and by Safety Evguardins dated Fncbyuans      25, 1004, Fbamney 24, 1008, ptubrsa8, 1008, Dotembie FRTh  Icese    22, 108, shall     sMay 4,A1000, fulimeen          and May and maintain  in e5, 1000, allffect     tethe subjeetprovisionsGfollowing pevdFe the     of The liCFns7.      may make2hangds to the appnOvcd Fitte Putho tity Prof0 aFR Withoult prior approval of the Commissien only ifthese ehanges would not E. The licensee shall fully implement and maintain in effect all provisions of the Comm ission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provision of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CIFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Florida Power and Light
                        & FPL Energy Seabrook Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan - Revision 3 "submitted by letter dated May 18, 2006.                                                                             0 The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Turkey Point Nuclear Generating Station CSP was approved by License Amendment No. 245.
F. 1. The licensee shall restrict the combined number of fuel assemblies loaded in the existing spent fuel pool storage racks and cask pit rack to no more than the capacity of the spent fuel pool storage racks. This condition applies at all times, except during activities associated with a reactor core offload/reload refueling condition. This restriction will ensure the capability to unload and remove the cask pit rack when cask loading operations are necessary.
: 2. The licensee shall establish two hold points within the rack installation procedure to ensure proper orientation of the cask rack in each unit's spent fuel pool. Verification of proper cask pit rack orientation will be implemented by an authorized Quality Control inspector during installation of the racks to ensure consistency with associated spent fuel pool criticality analysis assumptions.
FRenewed License No. DPR-31 Amendment No. 246 I    I   I    I    I

PTN Unit 4 Facility Operating License and Technical Specification mark-up LAR Attachment N 4
D lcftcrc dated April 3,10 80, July 0, 1080, Dcccmbcr 8,10 80, January 26, 1081, Ilnsert Attachment 2      May 10,1082, March 27,1084, April 16,1084, Au              1           and by Safety EvaluatiR6 diatcd- Februar-y 25, 10014, February 2110,Otober 8, 10089, Deeccmbcr 22, 10098, May 4,10090, and Mas,00 ubjcct to the following Thc liccnscc may makc chaRgcs tc thc approv-d Firc Prctcctcn PrOgra-m Withoult prior approval of the Cornmissien only if those changcz wouldl not advcrzcly affect the ability to aehievc and maintain safc shutdewn in the cece       o E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provision of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Florida Power and Light
                          & FPL Energy Seabrook Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan Revision 3, "submitted by letter dated May 18, 2006 The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Turkey Point Nuclear Generating Station CSP was approved by License Amendment No. 241.
F. 1. The licensee shall restrict the combined number of fuel assemblies loaded in the existing spent fuel pool storage racks and cask pit rack to no more than the capacity of the spent fuel pool storage racks. This condition applies at all times, except during activities associated with a reactor core offload/reload refueling condition. This restriction will ensure the capability to unload and remove the cask pit rack when cask loading operations are necessary.
: 2. The licensee shall establish two hold points within the rack installation procedure to ensure proper orientation of the cask rack in each unit's spent fuel pool. Verification of proper cask pit rack orientation will be implemented by an authorized Quality Control inspector during installation of the racks to ensure consistency with associated spent fuel pool criticality analysis assumptions.
FRenewed Limcense No., DPR-4 Amend ment No. 4

ADMINISTRATIVE CONTROLS 6.8   PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
: a.      The applicable procedures required by the Quality Assurance Topical Report.
: b.      The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; C.      Process Control Program implementation;
: d.      Offsite Dose Calculation Manual implementation;
: e.      Quality Control Program for effluent monitoring using the guidance in Regulatory Guide 1.21,
: f.      FlFP             c      Padd                  "Deleted"
: g.      Quality Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975; and
: h.      Diesel Fuel Oil Testing Program implementation.
6.8.2 Deleted 6.8.3 Deleted TURKEY POINT - UNITS 3 & 4                           6-13                                         236 AND 234 EAMENDMENT NOS.
FPL Turkey Point Nuclear Plant Unit 3 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated (and supplements dated _)and                 as approved in the safety evaluation report dated                 (and supplements dated     _).              Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1X10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Other Changes that May Be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.
Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
        " Fire Alarm and Detection Systems (Section 3.8);
        " Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
        " Gaseous Fire Suppression Systems (Section 3.10); and,
        " Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated                    to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Transition License Conditions
: 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2. The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S, Table S-2.
: 3. The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
FPL Turkey Point Nuclear Plant Unit 4 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated (and supplements dated                    )and as approved in the safety evaluation report dated                  (and supplements dated      _).             Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1X1 0 7/year (yr) for CDF and less than lxl0-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Other Changes that May Be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.
Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
       " Fire Alarm and Detection Systems (Section 3.8);
       " Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
       " Gaseous Fire Suppression Systems (Section 3.10); and,
       " Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated                    to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Transition License Conditions
: 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2. The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S, Table S-2.
: 3. The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.

Florida Power & Light              Attachment N - Technical Specification Changes FACILITY OPERATING LICENSE, TECHNICAL SPECIFICATIONS RETYPE 9 Pages Attached Page N-4 Revision 00 Revision                                                                  Page N-4

3 E.      Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F.      Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.
: 3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:
A.      Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).
B.      Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.        are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
C.      Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.
The licensee shall complete these activities no later than July 19, 2012.
The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Renewed License No. DPR-31 Amendment No.

4 D. Fire Protection FPL Turkey Point Nuclear Plant Unit 3 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated                    (and supplements dated _)                  and as approved in the safety evaluation report dated                      (and supplements dated                   ). Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a)      Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b)      Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 X10-7/year (yr) for CDF and less than lxl0-8 /yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Renewed License No. DPR-31 Amendment No.

5 Other Changes that May Be Made Without Prior NRC Approval
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.
Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
0        Fire Alarm and Detection Systems (Section 3.8);
0       Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
a        Gaseous Fire Suppression Systems (Section 3.10); and, E        Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Renewed License No. DPR-31 Amendment No.
0

6 Transition License Conditions
: 1.       Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2.       The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S, Table S-2.
: 3.       The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provision of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Florida Power and Light
   & FPL Energy Seabrook Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan - Revision 3 "submitted by letter dated May 18, 2006.
The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Turkey Point Nuclear Generating Station CSP was approved by License Amendment No. 245.
F. 1. The licensee shall restrict the combined number of fuel assemblies loaded in the existing spent fuel pool storage racks and cask pit rack to no more than the capacity of the spent fuel pool storage racks. This condition applies at all times, except during activities associated with a reactor core offload/reload refueling condition. This restriction will ensure the capability to unload and remove the cask pit rack when cask loading operations are necessary.
: 2. The licensee shall establish two hold points within the rack installation procedure to ensure proper orientation of the cask rack in each unit's spent fuel pool. Verification of proper cask pit rack orientation will be implemented by an authorized Quality Control inspector during installation of the racks to ensure consistency with associated spent fuel pool criticality analysis assumptions.
Renewed License No. DPR-31 Amendment No.

3 E.      Pursuant to the Act and 10 CFR Parts 40 and 70 to receive, possess, and use at any time 100 milligrams each of any source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactively contaminated apparatus; F.      Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of Turkey Point Units Nos. 3 and 4.
: 3. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:
A.      Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2300 megawatts (thermal).
B.      Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.        are hereby incorporated into this renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
C.      Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.
The licensee shall complete these activities no later than April 10, 2013.
The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Renewed License No. DPR-41 Amendment No.

4 D. Fire Protection FPL Turkey Point Nuclear Plant Unit 4 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated                     (and supplements dated _)                 and as approved in the safety evaluation report dated                       (and supplements dated   _).                 Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a)      Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b)      Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 X10- /year (yr) for CDF and less than 1 x1 0 8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Renewed License No. DPR-41 Amendment No.

5 Other Changes that May Be Made Without Prior NRC Approval                         i
: 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.
Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
0       Fire Alarm and Detection Systems (Section 3.8);
0       Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);
0       Gaseous Fire Suppression Systems (Section 3.10); and, 0       Passive Fire Protection Features (Section 3.11).
: 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation report dated to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Renewed License No. DPR-41 Amendment No.
0

6 Transition License Conditions
: 1.       Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
: 2.       The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S, Table S-2.
: 3.       The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
E. The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provision of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Florida Power and Light
   & FPL Energy Seabrook Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan Revision 3, "submitted by letter dated May 18, 2006 The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The Turkey Point Nuclear Generating Station CSP was approved by License Amendment No. 241.
F. 1. The licensee shall restrict the combined number of fuel assemblies loaded in the existing spent fuel pool storage racks and cask pit rack to no more than the capacity of the spent fuel pool storage racks. This condition applies at all times, except during activities associated with a reactor core offload/reload refueling condition. This restriction will ensure the capability to unload and remove the cask pit rack when cask loading operations are necessary.
: 2. The licensee shall establish two hold points within the rack installation procedure to ensure proper orientation of the cask rack in each unit's spent fuel pool. Verification of proper cask pit rack orientation will be implemented by an authorized Quality Control inspector during installation of the racks to ensure consistency with associated spent fuel pool criticality analysis assumptions.
Renewed License No. DPR-41 Amendment No.

ADMINISTRATIVE CONTROLS 6.8   PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:
: a.      The applicable procedures required by the Quality Assurance Topical Report.
: b.      The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; C.      Process Control Program implementation;
: d.      Offsite Dose Calculation Manual implementation;
: e.      Quality Control Program for effluent monitoring using the guidance in Regulatory Guide 1.21, Revision 1, June 1974;
: f.      Deleted
: g.      Quality Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975; and
: h.      Diesel Fuel Oil Testing Program implementation.
6.8.2 Deleted                                                                                                  0 6.8.3 Deleted TURKEY POINT - UNITS 3 & 4                           6-13                          AMENDMENT NOS.          AND

Florida Power & Light                       Attachment 0 - Orders and Exemption Florida Power & Light                       Attachment 0- Orders and  Exemption
: 0. Orders and Exemptions 3 Pages Attached Page 0-1 Revision 00 Revision                                                                Page 0-1

Florida Power & Light                                  Attachment 0 - Orders and Exemption Exemptions Rescind the following exemptions granted against 10 CFR 50, Appendix R:
    " LA-01-19840327, 10 CFR 50, Appendix R Exemption from Fixed Fire Suppression in the Control Room (III.G.3 Criteria)
    " LA-02-19840327, 10 CFR 50, Appendix R Exemption from Having an Oil Collection Tank Sized to Contain the Entire Lube Oil System Inventory (111.0 Criteria)
    " LA-03-19840327, 10 CFR 50, Appendix R Exemption from Providing 3-Hour Rated Fire Barriers Between Fire Zones 11, 12, 13, 14, 15 and 16 (lII.G.2.a Criteria)
    " LA-04-19840327, 10 CFR 50, Appendix R Exemption from Providing a 3-Hour Rated Fire Door for the Charging Pump Rooms on the West Walls of Fire Areas N and 0 (IllI.G.2.a Criteria)
    " LA-05-19840327, 10 CFR 50, Appendix R Exemption from Providing a Fire Suppression System for the Charging Pump Rooms Fire Areas N and 0 (IIl.G.2.c Criteria)
     " LA-06-19840327, 10 CFR 50, Appendix R Exemption from Providing Emergency Lighting Units in the Units 3 and 4 Containment to Facilitate the Manual Operation of Accumulator Stop Valves and RHR Inlet Isolation Valves (lIl.J Criteria)
     " LA-07-19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria)
    " LA-08-19840327, 10 CFR 50, Appendix R Exemption from Providing the Enclosure of One Charging Pump in a 1-Hour Barrier for Fire Areas N and 0 (llI.G.2.c Criteria)
     " LA-09-19840327, 10 CFR 50, Appendix R Exemption from Total Enclosure of One Diesel Generator Radiator Room By a 3-Hour Rated Fire Barrier (llI.G.2.a Criteria)
     " LA-10-19870812, 10 CFR 50, Appendix R Exemption for Intervening Combustibles Inside the Primary Containment (III.G.2.d Criteria)
     " LA-11-19870812, 10 CFR 50, Appendix R Exemption from Detection and Automatic Fire Suppression Systems in Outdoor Fire Areas 76, 77, 78, 83, 87, and 90 which are Separated by a Horizontal Distance of 20 Feet or Greater (llI.G.2.b Criteria)
     " LA-12-19870812, 10 CFR 50, Appendix R Exemption from Providing Detection and Automatic Fire Suppression in Outdoor Fire Zones 76, 77, 78, 83, 87, and 90 which are Enclosed in a Fire Barrier Having a 1-Hour Rating (llI.G.2.c Criteria)
     " LA-13-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area AAA from Fire Area A (IllI.G.2.a Criteria)
Page 0-2 Revision 00                                                                         Page 0-2

Florida Power & Liaht                                      Attachment 0 - Orders and Exemption
    "   LA-14-19870812, 10 CFR 50, Appendix R Exemption from Separation of Cables and Equipment and Associated Non-Safety Circuits or Redundant Trains by a Fire Barrier Having a 3-Hour Rating between Fire Area F from Fire Area A (lll.G.2.a Criteria)
    "   LA-15-19980224, 10 CFR 50, Appendix R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier Until a Horizontal Distance of 10 Feet is Attained in Roof Top Locations Excluding the Turbine Building (llI.G.2.a Criteria) a   LA-16-19980224, 10 CFR 50, Appendix R Exemption from IIl.G.2.a for 3 Hr Electrical Raceway Fire Barriers in the Outdoor Fire Zones 47, 54, 113, 114, 115, 116, 119, and 120 Excluding the Turbine Building (llI.G.2.a Criteria)
    "   LA-17-19981008, App R Exemption from IIl.G.2.a by a 25-Minute Rated Fire Barrier until a Horizontal Distance of 20 Ft is Attained in the Outdoor Fire Zones 79-partial, 84-partial, 86, 88-partial, and 89-partial Excluding the Turbine Bldg (lll.G.2.a Criteria)
     "  LA-18-19981008, 10 CFR 50, Appendix R Exemption from IIl.G.2.a West of the Open Turbine Building Structure Column Line A by a 1-Hour Rated Fire Barrier Until a Horizontal Distance of 20 Feet is Attained in Fire Zones 81 and 86 (llI.G.2.a Criteria)
    "   LA-19-19990504, 10 CFR 50, Appendix R Exemption from Technical Requirements for Outdoor Fire Zone 106R to the Extent that it had not Protected a Redundant Train with a 3-Hour Rated Barrier (llI.G.2.a Criteria)
     "  LA-20-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using a Minimum 1-Hour Barrier between Column Lines A and E-1(III.G.2.a Criteria)
    "   LA-21-19990505, 10 CFR 50, Appendix R Exemption for Raceway Fire Barriers in the Open Turbine Building from IIl.G.2.a Using Separation by a 25-Minute Fire Rated Barrier between Column Lines E-1 and Jc and the Turbine Deck (llI.G.2.a Criteria)
    "   LA-22-20060927, 10 CFR 50, Appendix R Exemption for Lack of Suppression in the Mechanical Equipment Room (Fire Zone 097) and for Lack of Suppression and Detection in the Control Room Roof (Fire Zone 106R) (III.G.3 Criteria)
Specific details regarding these exemptions are contained in Attachment K.
Orders No Orders need to be superseded or revised.
PTN implemented the following process for making this determination:
A review was conducted of the PTN docketed correspondence by PTN licensing staff. The review was performed by reviewing the correspondence files and performing electronic searches of internal PTN records and the NRC's ADAMS document system.
A specific review was performed of the license amendment that incorporated the mitigation strategies required by Section B.5.b of Commission Order EA-02-026 (TAC No's MD4560 and MD4561) to ensure that any changes being made to ensure Revision 0                                                                            Page 0-3

Florida Power & Light                                 Attachment 0 - Orders and Exemption compliance with 10 CFR 50.48(c) do not invalidate existing commitments applicable to the plant. The review of this order demonstrated that changes to the fire protection program will not affect measures required by B.5.b.
0 Page 0-4 Revision 00 Revision                                                                          Page 0-4

Florida Power & Light       Attachment P - RI-PB Alternatives to NFPA 806 10 CFR 60.48(c)(4)
P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)
No risk-informed or performance-based alternatives to compliance with NFPA 805 (per 10 CFR 50.48(c)(4)) were utilized by PTN.
Page P-I Revision 00 Revision                                                                          Page P-1

Florida Power & Light                 Attachment Q - No Significant Hazards Evaluation Q. No Significant Hazards Evaluations 3 Pages Attached Revision 0                                                                   Page Q-1

Florida Power & Light                         Attachment Q - No Siqnificant Hazards Evaluation FPL has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
: 1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated?
Response: No.
Operation of PTN in accordance with the proposed amendment does not increase the probability or consequences of accidents previously evaluated. The Updated Final Safety Analysis Report (UFSAR) documents the analyses of design basis accidents (DBAs) at PTN. The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility and does not adversely affect the ability of structures, systems, and components (SSCs) to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown (SSD) condition will remain capable of performing their design functions.
The purpose of this amendment is to permit PTN to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205. The NRC considers that National Fire Protection Association (NFPA) 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). Engineering analyses, in accordance with NFPA 805, have been performed to demonstrate that the risk-informed, performance-based (RI-PB) requirements per NFPA 805 have been met.
NFPA 805, taken as a whole, provides an acceptable alternative to 10 CFR 50.48(b) and satisfies 10 CFR 50.48(a) and General Design Criterion 3 of Appendix A to 10 CFR 50 and meets the underlying intent of the NRC's existing fire protection regulations and guidance, achieves defense-in-depth (DID) and the goals, performance objectives, and performance criteria specified in Chapter 1 of the standard. The small increase in net change in core damage frequency associated with this License Amendment Request (LAR) submittal is consistent with the Commission's Safety Goal Policy. Additionally, 10 CFR 50.48(c) allows self approval of fire protection program changes post-transition. If there are any increases post-transition in core damage frequency or risk, the increase will be small and consistent with the intent of the Commission's Safety Goal Policy.
Based on this, the implementation of this amendment does not significantly increase the probability of any accident previously evaluated. Equipment required to mitigate an accident remains capable of performing the assumed function.. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of this amendment.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any kind of accident previously evaluated?
Revision 0                                                                          Page Q-2

Florida Power & Light                          Attachment Q - No Significant Hazards Evaluation Response: No.                                                                               0 Operation of PTN in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Any scenario or previously analyzed accident with offsite dose was included in the evaluation of DBAs documented in the UFSAR. The proposed change does not alter the requirements or function for systems required during accident conditions. Implementation of the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205 will not result in new or different accidents.
The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition remain capable of performing their design functions.
The purpose of this amendment is to permit PTN to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004).
The requirements in NFPA 805 address only fire protection and the impacts of fire on the plant that have already been evaluated. Based on this, the implementation of this amendment does not create the possibility of a new or different kind of accident from any kind of accident previously evaluated. The proposed changes do not involve new failure mechanisms or malfunctions that can initiate a new accident.
Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created with the implementation of this amendment.
: 3. Does the proposed amendment involve a significant reduction in the margin of safety?
Response: No Operation of PTN in accordance with the proposed amendment does not involve a significant reduction in the margin of safety. The proposed amendment does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the UFSAR. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition remain capable of performing their design function.
The purpose of this amendment is to permit PTN to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) 0 Revision 0                                                                            Page Q-3

Florida Power & Light                          Attachment Q - No Significant Hazards Evaluation and the guidance in Revision 1 of RG 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004).
Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based methods do not result in a significant reduction in the margin of safety.
Based on this, the implementation of this amendment does not significantly reduce the margin of safety. The proposed changes are evaluated to ensure that the risk and safety margins are kept within acceptable limits. Therefore, the transition does not involve a significant reduction in the margin of safety.
NFPA 805 continues to protect public health and safety and the common defense and security because the overall approach of NFPA 805 is consistent with the key principles for evaluating license basis changes, as described in RG 1.174, is consistent with the DID philosophy, and maintains sufficient safety margins.
Margins previously established for the PTN program in accordance with 10 CFR 50.48(b) and Appendix R to 10 CFR 50 are not significantly reduced. Therefore, this LAR does not result in a reduction in a margin of safety.
Based on the above, FPL concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and accordingly, a finding of no significant hazards is justified.
In conclusion, 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.
Page Q-4 Revision 00                                                                           Page Q-4

Florida Power & Light                       Attachment R -Environmental Considerations R. Environmental Considerations Evaluation I Page Attached Page R-1 Revision 00                                                                  Page R-1

Florida Power & Light                                Attachment R -Environmental Considerations FPL has evaluated this License Amendment Request (LAR) against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. FPL has determined that this LAR meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50.
The purpose of this amendment is to permit PTN to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205. The NRC considers that National Fire Protection Association (NFPA) 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004).
The requirements in NFPA 805 address only fire protection and the impacts of fire on the plant have already been evaluated, as part of compliance to 10 CFR 50.48(a) and (b).
This amendment meets the following specific criteria:
: i. The amendment involves no significant hazards consideration.
As stated in Section 5.3.1, this proposed amendment does not involve significant hazards consideration.
ii. There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for radioactive releases to the environment.
Transition to the NFPA 805 requirements does not impact effluents. Therefore, there will be no significant change in the types or significant increase in the amounts of any effluents released offsite.
iii. There is no significant increase in individual or cumulative occupational radiation exposure.
Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for occupational exposure. There will be no significant increase in individual or cumulative occupational radiation exposure resulting from this change.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in conjunction with the proposed amendment.
Page R-2 Revision 00                                                                             Page R-2

Florida Power & Light                 Affachment T - Clarification of Prior NRC Approvals T. Clarification of Prior NRC Approvals 3 Pages Attached Page T-1 Revision 00                                                                      Page T-1

Florida Power & Light                          Aftachment T - Clarification of Prior NRC Approvals Introduction The elements of the pre-transition fire protection program licensing basis for which specific NRC previous approval is uncertain are included in this attachment. Also included is sufficient detail to demonstrate how those elements of the pre-transition fire protection program licensing basis meet the requirements in 10 CFR 50.48(c) (RG 1.205, Revision 1, Regulatory Position 2.2.1).
Prior Approval Clarification Request 1 Pre-transition Fire Protection Program Licensing Basis:
The PTN pre-transition licensing basis for the LA-07-19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria) was granted by an NRC SER dated March 27, 1984. The SER basis, however, used the term "free of intervening combustibles" which was not the intent of the FPL submittal provided.
Background/Basis:
FPL-NRC, L-83-347, 10 CFR 50 Appendix R Fire Protection, dated June 6, 1983, Fire Areas P & Q - Exemption Request P-1 and Q-1 The FPL submittal stated the following with the relevant words underlined for review:
     "Redundantequipment associated with operation of the pressurizer,including the pressurizerheaters, PORV's, Block Valves and Level Transmitters are located on or within the pressurizermissile shield wall as dictatedby plant design criteria. This area is void of significant insitu fsic, recte in situ7 combustibles, and during plant operation, consideredinaccessible to transient combustibles. Cable to this equipment is routed in a well separatedmanner to the maximum extent possible and rapidly achieves 20' separationas it leaves the area of the pressurizer.
Chargingline isolation valves are located within the biologicalshield wall in an area having no insitu [sic, recte in situ7 combustibles in proximity to the valves. Cables to these redundant valves are routed in steel conduit in the area of the valves and following proposed modifications, will rapidly achieve 20' separationas they leave the area of the valves.
Excess Letdown Isolation Valves are located within the biologicalshield wall in an area having no insitu fsic, recte in situ7 combustibles in proximity to the valves.
Cables to these redundant valves are routed in steel conduit in the area of the valves and following proposed modifications, will rapidly achieve 20' separationas they leave the area of the valves."
     "In conclusion, it is FPL's position that the proposed and existing fire protection features for Fire Area P provide a level of fire protection consistent with the fire hazards,both insitu fsic, recte in situ7 and transient,identified for this fire area and that these features provide a high level of assurance that at least one train of safe shutdown equipment and cables will remain free of fire damage."
     "In conclusion, it is FPL's position that the proposed and existing fire protection features for the Fire Area Q provide a level of fire protection consistent with the fire Reviion                                                                  Pag TI Revision 0                                                                                Page T-2

Florida Power & Light                            Attachment T - Clarification of Prior NRC Approvals hazards, both insitu [sic, recte in situl and transient identified for this fire area and that these features provide a high level of assurancethat at least one train of safe shutdown equipment and cables will remain free of fire damage."
NRC-FPL, Exemption Requests for Turkey Point Plant Units 3 and 4, dated March 27, 1984, SER Section 8.3 The NRC SER basis, however, did not provide the same basis details as the FPL submittal. It stated:
     "Section II1. G calls for installation of a radiantenergy shield between redundantsafe shutdown equipment and cables that are closer to each other than 20 feet horizontal distance. The licensee has proposed to eitherreroute redundantcables and equipment to provide 20 feet of separation,free of intervening combustibles, or provide 1-hour fire rated barriersand radiantenergy shields on all redundant cables in both units except for the redundant equipment associated with operation of the pressurizer,including the pressurizerheaters, PORVs, block valves and level transmitterslocated on or within the pressurizermissile shield walls. These areas are void of in-situ combustibles and are inaccessibleduring plant operation. All cabling inside the missile shield walls are routed in conduit."
FPL Licensing Interpretation of the NRC SER The NRC SER did not "repeat back" all of the details of the docketed PTN licensing and design basis. However, this was not uncommon for fire protection related SERs of that period. Implicit acceptance of the docketed licensing and design basis is assumed.
Request As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as a "prior approval" recognition that FPL did not and should not use the expression "free of intervening combustibles" in our text, however, the NRC used it.
As a result, this is being used to clarify the meaning of "free of intervening combustibles" in the March 27, 1984 SER as part of the licensing basis for the LA 19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria).
Prior Approval Clarification Request 2 Pre-transition. Fire Protection Program Licensing Basis:
The PTN pre-transition licensing basis for the LA-07-19840327, 10 CFR 50 Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lll.G.2.f Criteria) was granted by an NRC SER dated March 27, 1984. The NRC says in the SER that four portable fire extinguishers are available inside containment for use by the fire brigade which are currently not available as described since extinguishers are removed from the containment during operation.
Page T-3 Revision 00                                                                                 Page T-3

Florida Power & Light                        Attachment T - Clarification of Prior NRC Approvals Background/Basis:                                                                                0 FPL-NRC, L-83-347, 10 CFR 50 Appendix R Fire Protection, dated June 6, 1983, Fire Areas P & Q - Exemption Request P-1 and Q-1 The FPL submittal for the March 27, 1984 SER does not mention the fire extinguishers, just passive and active features.
     "Fireprotection features, existing and proposed for this fire zone are both active and passive in nature. Passive protection is provided for the redundant equipment and instrumentation and their associatedcables by existing and proposed physical separation and one-hour ratedprotection. Active fire protection is provided by the existing fire detection systems in the electrical penetrationarea which will provide rapid early warning of fire before significantdevelopment occurs."
NRC-FPL, Exemption Requests for Turkey Point Plant Units 3 and 4, dated March 27, 1984, SER Section 8.2 The NRC SER Discussion Section provides the following details:
     "No hose stations are located inside containment, however, four portable fire extinguishers are available inside containment for use by the fire brigade."
PTN-FPER-08-007, Attachment 2, dated March 11, 2011, Validation Report, LA  19840327 The validation report stated the following changes that have been made to the portable fire extinguishers in the Containment.
     "Portableextinguishers are no longerstaged inside containment. To avoid potential concerns from elevated temperatures, water extinguishers are staged inside during refueling outages and removed during operation. In addition, dry chemical extinguishers are mounted outside on the access stair handrailsand readily accessible for brigade use."
FPL Licensing Interpretation of the NRC SER The NRC SER does not reflect the current conditions and details of the docketed PTN licensing and design basis. Implicit acceptance of the docketed licensing and PTN design basis change is assumed.
Request As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC formally document as a "prior approval" recognition that portable extinguishers are no longer staged inside Containment. To avoid potential concerns from elevated temperatures, water extinguishers are staged inside during refueling outages and removed during operation. In addition, dry chemical extinguishers are mounted outside on the access stair handrails and readily accessible for brigade use. This is now part of the Discussion Section of LA-07-19840327, 10 CFR 50, Appendix R Exemption From Providing Non-Combustible Radiant Energy Shields in Fire Areas P and Q where Separation is Less than 20 Feet (lII.G.2.f Criteria).
0 Revision 0                                                                             Page T-4

Florida Power & Light                       Attachment U - Internal Events PRA Quality U. Internal Events PRA Quality 27 Pages Attached Page U-I Revision 00 Revision                                                                     Page U-1

Florida Power & LiQht                                        Attachment U - Internal Events PRA Quality The Turkey Point Internal Events Peer Review was performed in January 2002 using the NEI 00-02' process. Following the issuance of the ASME PRA Standard and Regulatory Guide 1.200 (RG 1.200), an internal gap analysis was performed where the findings from the original peer review were incorporated into the overall assessment of the PRA's quality with respect to the Standard's supporting requirements. The current PTN gap analysis uses the RA-Sa-2009 version of the standard as endorsed by RG 1.200, Revision 2.
To supplement the original peer review and internal gap analysis, and to further verify the quality of the updated internal events model used in the Fire PRA, in April 2011, focused peer reviews were performed for the human reliability analysis (HRA) and internal flooding analysis portions of the most current combined PRA standard, ASME/ANS RA-Sa-2009, and Regulatory Guide 1.200, Rev. 2. The internal flooding analysis focused peer review was performed because the latest internal flooding analysis was much more comprehensive than the original screening analysis that was performed for the IPEEE. Although the basic methods used for the HRA had not changed substantially, the HRA focused peer review was performed because of the enhanced HRA dependency analysis and the use of the HRA Calculator software in the latest model, and the fact that HRA plays a significant role in the determination of the dominant sequences and overall risk profile. These peer review reports are available upon request.
Table U-1 provides a list of the Facts and Observations (F&Os) from the original peer review and the focused peer reviews.
7 NEI-00-02, "Industry PRA Peer Review Process," Nuclear Energy Institute, January 2000.
Revision 0                                                                                   Page U-2

Florida Power & Light                                                                                                                     Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance             Possible Resolution        Plant Response or Resolution            SR    Impact on Fire PRA AS-1        The following items were observed related to     Given the large contribution to (1) Perform a more explicit  (1) For the small-small LOCA         AS-Al  F&O fully the success criteria:                            core damage frequency from      evaluation of the RWST       sequences where RWST refill is              addressed. No (1) For small - small LOCAs where high          small-small LOCA, with many      refill function to document  credited, secondary cooling is              impact.
pressure recirculation fails, the ECA-1.1 (Loss of the top cutsets related to    the ability to achieve a     available. Therefore, of ECC recirculation) actions to refill the     failure of the RWST refill       stable state end state. depressurization will take place and RWST via the CVCS and continue injection        function, a change in the        (2) Consider revising the   the leak rate reduced such that are modeled. Ifthis succeeds, the modeled       success criteria related to this charging pump modeling       RWST refill is viable. In most of endstate is successful core cooling,             function could have a            to be consistent with plant these sequences, some fault(s) in Although crediting the action is valid, the      significant impact on core       procedures.                 the RHR system is preventing sequence as modeled has not necessarily          damage frequency. The impact                                 successful recirculation or shutdown reached a stable end state; additional action   of the conservative modeling of                              cooling. While RWST refill may not inthe long term is required to put the plant in the charging pumps for the                                   be considered by some as a stable a stable state. For example, the RWST refill     emergency boration function                                  state, the small flow rate associated can be argued to extend the accident             was evaluated during the                                     with a depressurized small-small sequence mission time past 24 hours and          review. It was determined that                               LOCA and successful RWST refill therefore beyond the current Level 1 PRA         a change in the modeling                                     makes itfairly stable regardless.
model scope. Ifthe additional time were long,    would reduce core damage                                     The concern over extending the then in taking credit for these strategies the   frequency by approximately                                   mission time beyond 24 hours is impact on pump and other component run           1%.                                                          offset by the increasing probability failures, and any additional actions to achieve                                                               of recovering hardware failures in a stable state should be modeled. In addition,                                                                the cutset which prevented initiating some evaluation should be included                                                                            recirculation or shutdown cooling.
regarding potential effects on containment                                                                    Further, if RWST refill fails, HHSI instrumentation or components of increasing                                                                   from the opposite-unit pumps and water level.                                                                                                  RWST is available.
(2) For ATWS sequences where "Reactivity Control Late" is asked, the model credits emergency boration, per procedure FR-S.1.                                                                     (2) The charging pump success The procedure, and plant training documents,                                                                  criterion has been changed to 1/3 indicate that 1 of 3 charging pumps are                                                                       charging pumps for emergency needed to ensure at least 60 gpm of borated                                                                   boration.
injection for shutdown. However, the fault tree model (at gate U3WRCL1) implements this as failure if any single charging pump fails. This is incorrect and should be fixed.
AS-2       Several inconsistencies between the success      The success criteria             Revise the Accident         The updated Accident Sequence         SC-A3  F&O fully criteria as stated in the Accident Sequence      implemented inthe linked fault   Sequence Analysis           Analysis Notebook, the new                   addressed. No Analysis Notebook and the linked fault tree      tree model appear to be          Notebook and/or the         Success Criteria Calculation (PTN-           impact.
model. Specific examples are:                    adequate (and even               linked fault tree model to  BFJR-08-014), and the updated
: a. The Small-small LOCA success criteria for     conservative inthe case of the   ensure consistency          Revision 9 PTN PRA model (PTN-early core heat removal is listed in Table 3 as  AFW modeling for ATWS).          between the                 BFJR-00-001, Rev. 9) resolved the 2/4HHSI pumps and 1/3 AFW pumps.                 However, the inconsistencies     documentation and the       inconsistencies.
However, fault tree gate G1PMP3 shows 1          between the documentation        success criteria modeled HH pump required for small-small LOCA.           and the fault tree structure Revision 0                                                                                                                                                                    Page U-3

Florida Power & Light                                                                                                                 Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                           Significance            Possible Resolution       Plant Response or Resolution            SR    Impact on Fire PRA
: b. The Small LOCA success criteria for early    leads to confusion over which   in the linked fault tree.
core heat removal is shown as 2/4 HHSI          treatment is correct, creates pumps OR 1/2 RHR and depressurization.          questions concerning the However, section 4.1 of the Accident            thoroughness of independent Sequence Analysis Notebook and fault tree       review and may ultimately lead gate U3S2CD2 only credit HHSI.                  to questions about the validity
: c. The Medium LOCA success criteria for         of the model.
early core heat removal is listed as 2/4HHSI pumps inSection 5.1 and Table 5 of the Accident Sequence Analysis Notebook.
However, fault tree gate G1 PMP3 and the supporting MAAP analyses from the IPE show only 1 of 4 HH pumps to be required.
: d. The success criteria for early core heat removal using the AFW system is described differently in the PTN System Analysis Notebook and the Accident Sequence Analysis Notebook. The fault tree modeling appears to be generally consistent with the criteria stated in the System Analysis Notebook. For example, the System Analysis Notebook states that for ATWS, the AFW system must supply flow from 2 AFW pumps through all six AFW control valves. In Table 8 of the Accident Sequence Analysis Notebook, the ATWS success criteria for AFW is stated as 2 AFW pumps to 3/3 SGs. The structure of fault tree gate A0201 agrees with the System Notebook criteria rather than the Accident Sequence Analysis Notebook. Similar differences were noted in the success criteria descriptions related to the heat removal requirements for the Transient and Small-small LOCA sequences where the Accident Sequence Analysis only gives the pump success criteria without including the requirement to provide flow through at least 3 of 6 flow control valves.
AS-3       The Accident Sequence Analysis Notebook         The failure to credit alternate Consider removing non-     Continued HHSI using the opposite    SC-A3  F&O fully lists several systems available to provide      systems capable of satisfying   credited systems from the  unit RWST is now credited for long-         addressed. No various success criteria which are not          the functional success criteria discussion of systems      term cooling of transients. RWST            impact.
credited inthe linked fault tree. For example:  could lead to unnecessary       available to meet critical replenishment is not credited, based
: a. The Transient success criteria for long term conservatism inthe core         safety functions in the    on the judgment that the leak (2 core cooling does not credit RHR Shutdown       damage frequency results and    Accident Sequence          PORVs worth) is rather substantial, Revision 0                                                                                                                                                                Page U-4

0 Florida Power & Light                                                                                                                   Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                          Significance                Possible Resolution     Plant Response or Resolution           SR       Impact on Fire PRA Cooling, Low-head Recirculation, RWST         overestimation of the              Analysis Notebook or     and no secondary cooling is replenishment with continued HHSI,            importance of modeled              model the appropriate    available. The remaining options Depressurization with Low-head injection and  systems (e.g., the charging        alternate systems in the may be viable, but have many of the low-head recirculation, continued charging    pumps and boration system).        linked fault tree.       same hardware dependencies as with RWST replenishment, or continued HHSI     The decision to take little credit                          HHSR, and consequently will make with opposite Unit RWST.                       for the low-head system may                                 little difference and will not be
: b. The Small-small LOCA success criteria do   be based on the reliance of                                 added.
not credit Bleed and Feed and the long-term   these pumps on operator                                     As for small-small LOCA, continued cooling success criteria does not credit low- action to prevent overheating                               HHSI using the opposite unit RWST head recirculation or continued HHSI using    when operating above the                                    has been added. The addition of the opposite unit RWST                         shutoff head. This places                                   low-head recirculation with
: c. The Small LOCA success criteria for early   additional importance on                                    depressurization would have little core heat removal does not credit              resolving this issue by                                     effect due to shared dependencies depressurization and Low-head injection. In    procedure revision to address                               with HHSR and, therefore, will not addition, the long-term cooling function takes the requirement earlier or by                               be added. Credit for bleed-and-feed no credit for RHR cooling, low-head            additional analysis to increase                             cooling for the S1B sequences has recirculation or opposite unit RWST.           the time available for operator                             been added to the model.
response.                                                   For small LOCA, continued HHSI using the opposite unit RWST has been added. As for the addition of the other options, the addition of low-head recirculation with depressurization would have little effect due to shared dependencies with HHSR, and credit for depressurization and low-head injection was not modeled due to the reduced time available because of the larger break size (2"to 6").
The suggestion of removing mention of other, non-modeled alternatives from the AS notebook will not be taken. Removing them would only be removing information that may be useful in the future.
AS-9       The SGTR event tree does not have              This could have a noticeable       Ensure that there is     The SGTR event tree was revised     AS-Al     F&O fully complete sequence delineation. For the case    impact on LERF, and should         sufficient basis for the such that no credit for RCS                   addressed. No with failure of isolation and failure of HHSI, be considered / evaluated,         modeled accident         cooldown and depressurization is              impact. SGTR cooldown and depressurization to RHR                                              sequences.               given for sequences where SG                  not a part of fire conditions is asked and, ifsuccessful, no                                                                  isolation and HHSI fail.                      PRA model.
further delineation is provided in the model.
There should be some T/H analysis to show that the plant can be cooled down to RHR Revision 0                                                                                                                                                                  Page U-6

Florida Power & Light                                                                                                                    Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                       Description                               Significance            Possible Resolution       Plant Response or Resolution             SR    Impact on Fire PRA entry conditions without HHSI given a SGTR.
DA-1       NUREG/CR-4550 was followed to develop            CCF may significantly affect    Consider revising CCF       The Turkey Point CCF model was        DA-Cl  F&O fully CCF. NUREG/CR-4550 was issued in 1986.           the PRA results. A systematic   modeling based on the       updated to reflect the alpha-factor          addressed. No The approach addressed in NUREG/CR-4550          approach for CCF modeling is    NUREG/CR-4550               approach and data from INEL                  impact.
may be out of date. NUREG/CR-4780 (or            required for adequate analysis. systematic approach. A      94/0064.
equivalent) systematic approach should be                                        more updated systematic followed.                                                                        approach (such as NUREG/CR-4780 or equivalent) should be followed.
DA-2       The test and maintenance probabilities used      This is level B significance,   Consider revising the T &   Logic was introduced to the model     DA-C13 F&O fully for individual components are based on           because it would not have a     Mevent probabilities for    to change the opposite-unit EDG              addressed. No actual outage time as collected by the plant. significant effect on the risk  the opposite unit           test and maintenance probability             impact.
The component outage time was clearly            profile. The FV for EDG 4A/B is components to account       during outage conditions through collected over the period of time the plant was  about 2E-4, thus a low          for unavailability over the the use of flags representing the in Mode 1,2,3.                                   importance event. However, if   total period of demand. As  operating mode of the unit. These The fault trees and event trees use several      the CCF modeling included n-1   stated above, this can be   flags were also used to model the cross-ties from AC power, HHSI, and AFW. In      combinations, the importance    done at the fault logic     effect of the opposite unit's mode on the use of these cross-ties, the opposite unit   would be higher.                level or in the data        the different system crossties.
components have T&M events. The opposite                                         probabilities.
unit may be in Mode 4,5,6 at the time of demand and the desired equipment may have lesser Tech Specs than those assumed for power operation. The T&M event probabilities for the opposite unit components must consider unavailability over the total period of demand, not just during power operation. This can be done at the fault logic level (with house events for COS) or in the data probabilities. Currently, neither is done.
The most important case of this is the DGs.
The DG T&M unavailability is about 6E-3 (55 hours per year). Ifthe OOS time for major overhaul were considered, the unavailability would be .03 to .05.
DA-4       The latest data updating was done in 1995        The out of date plant specific  Review industrial events    The data used in the latest model     DA-C3  F&O fully based on plant specific data from 1990 to        data and generic database       after 1989 to update the    update has plant-specific data               addressed. No 1994, and the generic database developed in      may not reflect current plant   generic database and        derived from plant records from              impact.
1989.                                            conditions.                     update the reliability data 1992 through 2006, and generic based on current plant      data from the latest sources, specific data.              primarily NUREG/CR-6928.
Revision 0                                                                                                                                                                   Page U-6 0

Florida Power & Light                                                                                                                  Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                            Significance             Possible Resolution      Plant Response or Resolution            SR    Impact on Fire PRA DA-7        The following observations on CCF modeling     This is a level B because it                               For the first comment, it is         IE-C8   F&O fully in CCW and ICW:                                represents a potential                                     appropriate that the CCF for the             addressed. No For the ICW and CCW initiating events, RIF     underestimation of the ICW                                 ICW pumps be modeled this way for            impact.
of one pump to run for I year is coupled with  and CCW IEfrequencies.                                     the initiating event of loss of ICW.
CCF of the remaining 2 pumps to run for 72                                                                Two pumps are normally running hours, but the events are independent.                                                                    with one instandby. It makes no For lCW initiating event, CCF of the strainers                                                            sense to have a CCF event where are not included.                                                                                         all 3 pumps fail to run due to a There is no evidence of CCF caused by                                                                     common cause over a period of 1 environmental factors inthe ICW initiating                                                                year event.                                                                                                    CCF for the strainers was added to the model.
Review of PTN history showed no reactor trips due to environmental factors (sea grass). The initiating event fault tree for ICW includes CCF failures of the ICW pumps and strainers.
DE-1       Turkey Point switched their common cause        Failing to model the higher      As a minimum,             A complete CCF update addressed      DA-D5   F&O fully methodology from the Beta Factor Approach      order common cause sets (i.e. incorporate all higher    these issues. See the details of the         addressed. No to the Alpha Factor Approach. It appears that   those involving the failure of   order common cause sets   referenced PTN changes and the               impact.
Turkey Point only models the exact              more components than the         inthe model for all       PTN CCF calculation PTN-BFJR-combination of failures that lead to system     minimum required for system      components with an order  08-012, Rev. 0.
failure. For example, Gate G121 of the HPSI     failure can impact the CDF       of redundancy greater model covers the scenario where the success     because the probability of the   than 2. Turkey Point criterion is 2 of 4 HPSI pumps are required. higher order common cause        should also insert the However, only common cause failure of 3 of 4   sets may not be significantly    lower order common pumps is modeled. In this case common           less than the probability of the cause sets to ensure that cause failure of 4 of 4 pumps should also be    common cause failure of the      maintenance alignments modeled. Likewise, for gate G131, which         minimum set required for         are properly covered in covers the scenario where the success           system failure. Not modeling     the work week schedules.
criterion is 3 of 4 HPSI pumps, only common     the lower order sets probably cause failure of 2 of 4 pumps is modeled. In    will not affect the base case this case, common cause failure of 3 of 4       CDF but it could impact the pumps and 4 of 4 pumps should also be           results when evaluating the modeled. Turkey Point also does not model       CDF for specific maintenance common cause failure combinations less than     configurations where that required for direct system failure but     components are defined to be which could lead to system failure when         out of service.
combined with a single random failure. Even for the CCF level modeled, Turkey Point doesn't model all applicable combinations of CCF failures (i.e. for 2/4, there are six Revision 0                                                                                                                                                                 Page U-7

Florida Power & Light                                                                                                                       Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance               Possible Resolution        Plant Response or Resolution              SR      Impact on Fire PRA combinations). In addition, Turkey Point does not model common cause failure of the 2 Unit 3 EIDGs coupled with failure to align to the unit 4 EDGs.
DE-2       The Flooding analysis considered component       The estimated CDF is not          Review flooding analysis     The internal flooding analysis has     IFEV-A5 Internal flooding vulnerability to both flooding and spray         insignificant relative to the     to determine if the          been completely revised for RG                 is not effects. Multiple screening criteria were        current core damage               screening and flooding       1.200. See Calculation PTN-BFJR-               considered in employed to eliminate areas from further         frequency.                        initiator frequency          11-009, Rev. 0.                                the fire PRA.
consideration. The three remaining areas                                           calculations are still were then analyzed in more detail to                                               current using updated determine a CDF from flooding,                                                     methodology, models and A CDF of approximately 5E-7 was calculated                                         assumptions.
for flooding and determined to be not significant relative to then overall risk (IPE CDF of 3.7E-4). The current model maintains cutsets of lower CDF than these flooding cutsets. The conclusion that the flooding is not significant is no longer supported.
HR-A2-01   This HR requires identification, through a       A review of the HHSI system       A review of the              Rather than examine all possible       HR-D2   F&O fully review of procedures and practices, those        notebook identified 9             procedures listed in the     maintenance, surveillance, and                 addressed. No calibration activities that ifperformed          procedures inTable 5 that         system notebooks should      calibration procedures and                     impact.
incorrectly can have an adverse impact on        were associated with              be performed to identify     associated practices, a more the automatic initiation of standby safety       calibration of HHSI related       those that could result in   practical method was used which equipment. The system notebooks contain a        equipment, but NO                 potential miscalibration     presumed that pre-initiators can detailed listing of testing and maintenance      miscalibration events were        events, and provide a        potentially exist for all redundant procedures that were identified for each         listed inTable 7 - Pre-Initiator  justification for those that standby trains modeled in the PRA system, but there is no discussion as to which   Actions. The justification/basis  were excluded from           and to insert screening values for procedures were determined to have the           for exclusion of these            further consideration. For   their probability of occurrence. If potential to result in equipment being left in a procedures from resulting in a    miscalibrations that have    quantification of the model with the miscalibrated condition, and which were          miscalibration event could not    the potential to impact      screening values demonstrates that screened from consideration with the basis       be found. Additionally, a         multiple systems, ensure     they are risk significant contributors for screening.                                   process for                       that they are treated        (FV>0.005), then a specific review excluding/screening other         consistently between both    of potential maintenance, maintenance and testing           systems, and that            surveillance, and calibration procedures from being             appropriate HFEs are         procedures and practices that could potential pre-initiator events is listed in all impacted       cause the pre initiator condition to also not provided, resulting in   system notebooks.            exist is performed against the no traceability for how the                                    screening rules in HR B1. Any modeled pre-initiators were       Similar traceability needs   procedures that meet this criterion determined, and for how the       to be provided for other     are identified and documented in remaining procedures were         Test and Maintenance         the HRA Calculator file.
evaluated/screened from           procedures that have the     HR-Al presumes that EVERY further consideration.            ability to render a          possible procedure or practice that Revision 0                                                                                                                                                                         'Page U-8

Florida Power & Light                                                                                                                      Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                            Significance             Possible Resolution         Plant Response or Resolution              SR    Impact on Fire PRA system/equipment             could cause misalignment or In addition, a review of the     unavailable as well.         miscalibration are identified before HRA calculator files identified                               any screening of events can occur.
that at least one of the                                      Further, review of Table 7 revealed procedures referenced (3[4]-                                  that only 2 of the 9 calibration PMI-062.35 - RWST Level                                       procedures required pre-initiator Instrumentation Channel L                                  events for the PRA model, and the 6583A Calibration) was                                        events were already in the model.
modeled as a common cause miscalibration event in another system (RHR). The appropriate miscalibration event is in the fault tree model,, but is not identified or discussed inthe HHSI notebook, although the procedure resulting inthe miscalibration is identified.
HR-B2-01   This SR does not allow screening of activities Inthe HHSI system notebook,      Review the actual test and   For the 864 valves, the model has a     HR-B2  F&O fully that could simultaneously have an impact on    the following valves are         maintenance procedures       T&M event for each RWST to                     addressed. No multiple trains of a redundant system or       assumed not to be under          associated with these        account for the time the RWST                  impact.
diverse system.                                maintenance while either unit is valves and determine         contents are used to fill the refueling at power: MOV-*--,864A, B; *-    when they can be subject     canal, which is probably the only 864C; *-845A, B, C, D; MOV-      to testing or maintenance,   time the 864 valves could be 878A, B; MOV-* 856A, B;
* Ifthey can be subject to     maintained. The RWST T&M event 847C;
* 882. Because these       testing or maintenance       should be a palatable substitute valves have the potential to     when either of the Units is  The 845 and 882 valves are locked-impact BOTH Units, they          shutdown, then a T&M         open manual valves, so no T&M or cannot be screened inthis        needs to be added into       pre-initiator is needed there.
manner. Based on this            the model as well as assumption, these valves         consideration for a pre-     The HHSI recirculation valves 856 would only be worked on while    initiator misalignment of    and 874C, if closed for maintenance both Units are shutdown, which   the valves, and a post-      take out their related HHSI pumps.
is probably not realistic,       initiator HRA to re-align if The 856 valves are stroke-tested necessary.                   during the associated unit refueling outages. Evaluated pre-initiators for the 856 valves and added these to the model.
The 878A and 878B valves, if closed for maintenance, would prevent opposite-unit SI. Evaluated pre-initiators for the 878 valves and added these to the model.
Ran quantification of the U3 and U4 models with these pre-initiators Revision 0                                                                                                                                                                     Page U-9

Florida Power & Light                                                                                                                         Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                         Description                              Significance              Possible Resolution        Plant Response or Resolution               SR   Impact on Fire PRA added - negligible effect on CDF and LERF.
HR-C2-01   There is no provided documentation of the         CCII/Ill requires specific        Provide documentation of     In the latest data update,              HR-C2 F&O fully plant-specific or applicable generic operating    considerations from plant-        the review of plant-         documented in PTN-BFJR-02-026,                addressed. No experience for equipment left unavailable for     specific operating experience,    specific or generic          Rev. 1, condition reports were                impact.
response in accident sequences.                   The actual impact on the model    operating experience and     reviewed for the time period 1992-may be of little significance but confirm that no additional   2006 for component failures. No the lack of documentation on      failure mode is required. failure modes outside the ones plant-specific operating                                       already modeled were found.
experience makes the assessment not possible.
HR-D1-01   The human failure event probabilities appear      While the HRA calculator file     Review and ensure            The only thing that needs to be         HR-D1 Table in to be evaluated with a systematic process         and the CAFTA model are           consistency on the           fixed for this finding is Table 3 inthe       calculation will that includes an initial screening value and      consistent, there is no           calculation of HFE for all   09-11, Rev. 1 calc. The probabilities         be updated in the identification of risk-significant action for documentation on how the          pre-initiator and post-      for AHFAON2BK1,                               next revision.
which a detailed analysis through ASEP            actual value for the action is    initiator HFEs.              AHFAON2BK1U4, AHFAON2BK2,                     The model is method is used. Although there appear to be       calculated,                                                    and AHFAON2BK2U4 need to be                   correct. No some inconsistencies in the values of the                                                                        changed to 4.OE-05, matching the              impact.
HEF, especially for HEF already existing in                                                                      probabilities inTable 5 and the PRA previous version of the model. For example,                                                                      model.
action AHFAON2BK1 is indicated as a pre-                                                                         The probabilities for AHFAON2BKU existing action (i.e., not highlighted inTable 3,                                                                and AHFAON2BKUU4 need to be page 22) with an initial value of 1.10E-3.                                                                       changed to 3.6E-05, matching the There is no further discussion of this action                                                                    probabilities in Table 5 and the PRA (i.e., the action is not indicated in Table 4 at                                                                 model.
page 27 as one of the action requiring further analysis). Still inTable 5 at page 31 the                                                                        The manner inwhich these pre-action has a value of 4.5E-5 (consistently with                                                                  initiators are calculated is what is inthe model). Another example of                                                                         documented in the HRA Calculator inconsistency between the documentation,                                                                         file used and referenced in the HRA Calculator file and the CAFTA model                                                                      calculation PTN-BFJR-09-011, Rev.
is post-initiator action AHFPAFWTHROT).                                                                          1.
Only the tables, not the model, need to be changed.
HR-D3-01   The pre-initiator HRA does not specifically       HR-D3 requires, for each          To achieve CCII/III, a       For each pre-initiator analyzed in      HR-D3 F&O fully discuss quality of the written procedures or      detailed human error              discussion of the quality of detail in the HRA Calculator, there           addressed. No the quality of the human-machine interface,       probability assessment, a         procedures and human-        is an assessment of the quality of            impact.
discussion on the quality of      machine interface need to    the procedures and human interface procedures and human-             be provided,                 in Performance Shaping Factors.
machine interface.
Revision 0                                                                                                                                                                       Page U-10

0 Florida Power & Light                                                                                                                        Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                       Description                              Significance             Possible Resolution        Plant Response or Resolution               SR   Impact on Fire PRA HR-G7-01   This SR outlines the requirements for             A very detailed dependency       Review the cutsets           The fatigue rule has forced the site   HR-G7  F&O fully assessing the degree of dependence                assessment has been              associated with Dual-Unit    into a 4th RO on shift. Attachment 1          addressed. No between HEPs contained in a single accident       performed for Turkey Point.      initiating events, and take  of the Ops Dept Instruction ODI-              impact.
sequence or cutset, and accounting for the        However, one potential area      into consideration the fact  CO-045, Shift Staffing and influence of success or failure in preceding      that does not appear to have     that only 1 of the 2 Units   Accountabilities, 12/21/09 has to be human actions and system performance on           been fully addressed for HEPs    will have 2 Reactor          filled out by the Field Supervisor for the HEP under consideration, including            credited during dual-unit        Operators available, or      all of the bargaining unit operators consideration of 1) time required to complete     initiating events is the         that each Unit may have a    to ensure all of the required all actions in relation to the time available, 2) availability of Operator         single dedicated RO, and     positions are covered. Note that factors that could lead to dependence             resources inthe Control Room. a shared RO, and             there is a "FOURTH" RCO position including common instrumentation,                 All HEPs reviewed had "2         determine the impact of      that they have to fill now. That procedures, increased stress levels, etc., and    ROs" required under the          this resource limitation on  should give you what you need to
: 3) availability of resources (e.g. personnel). manpower requirements table. the HEP. This review         show 2 RCOs per unit on shift. The However, based on current        should include ALL dual      ADMs on the subject have been minimum Control Room             Unit initiating events that  changed as well to account for staffing, only 3 total ROs are   credit any post-initiator    assignment of the 4th RCO, but required to support both Units. actions, not only those      they have been carefully worded not Therefore, during a Dual Unit    with multiple actions since  to contradict Tech Specs which only initiating event there would not the Initiating Event itself  requires three... even though the be 2 ROs available to support    creates the dependency       attached ODI-CO-045 is what they both Units response times.       concern,                     actually use to make sure all of the shift assignments are covered.
IE-3       The initiator for Pressurized Thermal Shock is    This is level B because these    Provide justification for    Added RV Rupture IE in model           IE-A2  F&O fully considered "out of scope" with no justification. initiators have been identified  elimination of these         update. Research over the last few            addressed. No PTS is not out of scope for PSA. Discussions      inNUREG-1 150 as worthy of       initiators or include in     years by NRC, EPRI, DOE, MRP,                 impact.
with PTN PRA staff indicate that from a           consideration. There is no       model,                       and others has shown PTS to be a licensing perspective, it has been determined     basis for the elimination of                                  non-issue for plants like Turkey to be a resolved issue and therefore should       these initiators.                                             Point.
not be inthe PRA. This should be explained, with a probabilistic explanation of why it is a small contributor.
The initiator for Reactor Vessel Rupture is considered out of scope with no justification.
IE-4       PTN has several shared systems. These             This is a level B, because it    Develop and implement        See resolution of DA-2 for 1) and 3). DA-C13 F&O fully include HHSI, AFW, and DG. The                    could have significant effect on criteria for modeling dual   2) is addressed by PRA Change                 addressed. No configuration of these systems may depend         the availability of cross-tie    unit plant operating status. PTN-02-005, which is implemented              impact.
on the status of the unit. For the purpose of     systems or success criteria,                                  in the current PRA model.
system sharing, the opposite unit equipment is always assumed to be available in Mode 1 operability.
The following observations were made:
: 1) The T&M unavailability for DG implies about 50 hours a year OOS. This can not Revision 0                                                                                                                                                                     Page U-11

Florida Power & Light                                                                                                               Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                            Significance           Possible Resolution    Plant Response or Resolution              SR        Impact on Fire PRA include time for annual overhaul, (which is done at shutdown). When the opposite unit is in Mode 6, the DG tech spec is reduced to 1 DG. The PRA does not capture this dependency on plant status.
: 2) When Unit 4 is in Mode 5 or 6, the HHSI for unit 4 is cooled by CCW on unit 3. The PRA does not capture this.
: 3) there are no overall guidelines and criteria for treatment of the opposite unit's operability and its effect on equipment availability.
IE-7       The level of independent technical review of     This is a level B, because it  Provide evidence of an  There is now a comment and            QU-F1      F&O fully PRA changes is indeterminate. There are no       was not possible to determine  independent technical   resolution section inthe model                   addressed. No comment and resolution sheets for the PRA        if the PRA changes have        review. Document how    update calculations.                             impact.
modification process. There is only a signoff    actually had an independent    review was conducted sheet on the overall update package (which       review. Without a detailed     and what comments were contains several individual changes).            review, the analysis is        made during the review.
inadequate.
IE-9       Random reactor coolant pump seal failure         This is a level B because S1   Include random RCP seal A review of NUREG/CR-5750             IE-A2      F&O fully has not been included. From NUREG/CR-            LOCA is the most important     LOCA as an initiator,   revealed that the random seal                    addressed. No 5750, this event can be about 1E-3. The S1       contributor to risk.                                   LOCA frequency is 2.5E-3 per year,               impact.
frequency is derived from pipe rupture failures                                                         and is based on 2 events: a 1975 only. Itdoes not include component leakage,                                                             event at Robinson-2, and a 1980 RCP seal LOCA, or any other sources of                                                                  event at ANO-1. The incident at small leakage.                                                                                          ANO-1 can be discounted not only due to the fact that itwas 20 years ago and design changes have likely been made to preclude similar events, but also due to the fact that ANO-1 is a B&W plant with substantial differences in RCP and RCP seal design. The incident at Robinson-2 was discounted due to changes in seal design and seal-related operating practices and procedures implemented in the last 37 years.
IFPP-B3-   This SR requires that an uncertainty             No uncertainty assessment has  PTN should perform an   A new section titled "Assumptions     IF-IFPP-B3 Internal flooding 01         assessment be included in the                    been included inthe            uncertainty assessment  and Sources of Uncertainty" has                  is not documentation,                                   documentation,                 and document the        been added.                                      considered in assessment.                                                              the fire PRA.
Revision 0                                                                                                                                                                  Page U.12

Florida Power & Light                                                                                                                   Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                      Description                            Significance              Possible Resolution       Plant Response or Resolution              SR     Impact on Fire PRA IFQU-A1-   This SR states: For each flood scenario,       No documentation has been        A review of the accident   Added to section 4.2 of Calculation   IF-IFQU- Internal flooding 01         REVIEW the accident sequences for the          provided which identifies that a sequences and              PTN-BFJR-1 1-009, Rev. 0 - "as a      Al       is not associated plant-initiating event group to     review of the accident           quantification results     result of flooding. "Itshould be               considered in confirm applicability of the accident sequence sequences has been               should be documented,      noted that the accident sequences              the fire PRA.
model. Ifappropriate accident sequences do     performed. In particular, for                               defined inthe internal events model not exist, MODIFY sequences as necessary       pipe break events that are                                  were used to quantify internal to account for any unique flood-induced        MSLB scenarios that have the                                flooding scenarios. Each scenario scenarios and/or phenomena in accordance       potential to impact other plant                             description identifies the existing with the applicable requirements described in  equipment (e.g. TDAFW                                       initiating event to which it is paragraph 4.5.2.                               pumps), ensure that the MSLB                                mapped. No new sequences or fault accident sequence followed is                               tree models were required." Section appropriate. For Circulating                                4.2.3.1 specifically discusses main Water expansion joint failures,                             steam and feed line breaks as being ensure that loss of Condenser                               explicitly addressed in the PRA.
Vacuum sequences are                                        Add to end of Section 4.2.3.1 - "This followed, etc.                                              scenario is not considered further because it is already modeled in the PRA."
IFQU-A5-   No human failure event discussion is           No human failure event           The documentation of       No HFE discussion because no          IF-IFQU- Internal flooding 01         presented in the analysis.                     discussion is presented inthe    human failure events       credit was taken for mitigating       A5       is not analysis other than mention      included in the analysis   operator actions. Added to Section             considered in that such events have been       should be provided.        3.1.2, paragraph 6 -at specific times          the fire PRA.
included inthe analysis.         Additionally, the scenario are noted. "These times are noted specific impact on PSFs    only to highlight the differences in should be documented in    time pressures between flood the analysis.              scenarios. No credit is taken for operator actions that would mitigate any leak, spray, or rupture.
Subsequently, no human reliability analysis was necessary."
Deleted - "Subsequently, human reliability analysis was applied in quantifying certain accident scenarios when the early termination of the release is to be considered so as to estimate the probabilities that a release would not be terminated before certain damage occurred."
As for the HFEs from the internal events analysis, the documentation of the internal flooding HRA is included in the calculation PTN-Revision 0                                                                                                                                                                  Page U-13

Florida Power & Light                                                                                                                Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                      Description                        Significance               Possible Resolution       Plant Response or Resolution             SR      Impact on Fire PRA BFJR-11-010, Rev. 0, which is referenced in the internal flooding analysis calculation PTN-BFJR-1 I-009, Rev. 0.
IFQU-A7-   This SR states: PERFORM internal flood     The internal flooding analysis    The quantification process  The quantification is discussed in   IF-IFQU- Internal flooding 01         sequence quantification in accordance with has been quantified in            should either be            Section 4.3 of calculation PTN-      A7       is not the applicable requirements described in   accordance with internal          documented in the           BFJR-11-009, Rev. 0.                          considered in paragraph 4.5.8.                           events quantification             flooding analysis, or ifthe                                               the fire PRA.
requirements; however,            same process has been supporting documentation          used elsewhere, the should be provided which          flooding analysis should describes the process.            point to that process.
Additionally a review of the quantification should be documented.
IFSN-A11-  This SR states: For multi-unit sites with  No multi-unit impacts have        While PTN may not be        Dual-unit impacts are discussed      IF-IFSN- Internal flooding 01         shared systems or structures, INCLUDE      been quantified or discussed,     particularly vulnerable to  where applicable in individual flood All      is not multi-unit scenarios.                      Multi-Unit shared systems (e.g. multi-unit impacts, such    scenarios. The effects of these are           considered in TDAFW pumps and HHSI              impacts need to be          automatically accounted for                   the fire PRA.
pumps) need to be addressed.      discussed in the            because the Unit 3 and Unit 4 documentation. Ifthere      models are linked.
are no shared systems or structures, this needs to be explicitly stated in the analysis.
IFSN-A16-  This SR provides the criteria under which  No discussion of human            Documentation should be     No credit is taken for mitigating    IF-IFSN- Internal flooding 01         human mitigative actions can be credited. mitigative actions is provided in provided which details the  actions. The times at which various  A16      is not the flooding study. It appears    human action being          equipment fail in each scenario do            considered in that in some cases human          credited, and the basis for not imply an end to the scenario,             the fire PRA.
mitigative actions are credited,  why they are valid for the however even this is not clearly  scenarios under which stated. See for example the       they are credited.
following excerpt from 4.2.1.88 "A2000-gpm major flood rupture will cause flooding of the Unit 3 and Unit 4 RHR pump and heat exchanger rooms and the submergence damage to the RHR pumps in 2 hours." It is unclear as to why the flood only persists for 2 Revision 0                                                                                                                                                                Page U-14 0

0 Florida Power & Light                                                                                                                       Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                               Significance             Possible Resolution       Plant Response or Resolution              SR       Impact on Fire PRA hours.
IFSN-A2-   No identification of flood alarms or floor drains  For each flood area a            PTN should document         No credit taken for operator action  IF-IFSN-A2 Internal flooding 01         has been made in the flood analysis                combination of the discussion    and identify the presence   to mitigate flood; therefore, there             is not document,                                          inSection 4.2 and the            of flood alarms and floor   was no need to credit flood alarms.             considered in documentation found in            drains as related to their  Add to end of Section 3.1.3 - "In               the fire PRA.
Appendix B partially meets the    treatment in the analysis. looking at flood propagation by intent of this SR. Additional                                backflow through shared drain lines, information needs to be                                      no credit was taken for check addressed however specifically                                valves."
treatment of floor drains and flood alarms. No indication was                               Drain lines were not credited in given as to what the                                          determining the impact of a flood in justification to when a potential                             a particular room.
flood source is isolated                                      Also not credited are the level (possible impact due to flood                                 alarms for the RHR Room Sumps alarms) and no discussion has                                 and Switchgear Room Sump Pump been provided regarding                                       start alarms.
treatment of backflow through floor drains.
IFSN-A3-   This SR states: for each defined flood area       No justification for how/when     PTN should identify         No credit is taken for mitigating    IF-IFSN-A3 Internal flooding 01         and each flood source, IDENTIFY those             Operators determine a flood is    responses which have the    actions.                                        is not automatic or operator responses that have         in progress, and what triggers    ability to terminate or                                                     considered in the ability to terminate or contain the flood     them to attempt to isolate it can contain flood propagation,                                                  the fire PRA.
propagation.                                      be found. The front part of the   and provide a justification document states that all floods   for the timing used in the are assumed to last 12 to 24      analysis hours, but a review of the scenarios shows that some are terminated in 2 hours, others in 3 hours, etc. The basis for these times cannot be found.
No operator actions regarding termination have been addressed in the analysis.
Section 3.1.2 alludes to the fact that floods were generally allowed to persist for 12-24 hours, however discussion found in Section 4.2 seems to contradict this. No information has been provided detailing when floods have been isolated if credit was taken for Revision 0                                                                                                                                                                        Page U-15

Florida Power & Light                                                                                                                     Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance             Possible Resolution       Plant Response or Resolution             SR       Impact on Fire PRA such an action.
IFSN-A4-    No supportinginformation has been provided      No supporting information has    PTN should document the     The flooding calculations will be   IF-IFSN-A4 Internal flooding 01         to justify the estimations regarding flood       been provided to justify the     calculations performed in   added to the notebook along with               is not volumes and the subsequent flooding height.      estimations regarding flood      determining flood volumes   added discussion inSection 3.2.                considered in volumes and the subsequent       in a given flood area as it                                                the fire PRA.
flooding height.                 relates to equipment in the room (the floor area the equipment takes up),
the capacity of the system, the length of time the flood persists, etc.
IFSN-A6-   This SR States: For the SSCs identified in IF-   No discussion has been           Analysis should be          Added to section 3.1.2, paragraph 5 IF-IFSN-A6 Internal flooding 01         C2c, IDENTIFY the susceptibility of each         provided for the impact due to   performed which includes    end - "Inlight of this, itshould be            is not SSC in a flood area to flood-induced failure     the additional flood failure     failure by submergence or   noted that only spray and                      considered in mechanisms.                                      mechanisms. The analysis         spray, and a qualitative    submergence damage were                        the fire PRA.
INCLUDE failure by submergence and spray         refers to the fact that HELB is  assessment of other         included inthe scope of this in the identification process.                   not a concern at PTN but the     failure mechanisms needs    evaluation."
EITHER:                                         other failure mechanisms are     to be provided (e.g. jet a) ASSESS qualitatively the impact of flood-     not discussed.                   impingement, pipe whip, induced mechanisms that are not formally                                          humidity, condensation, addressed (e.g., using the mechanisms listed                                      temperature concerns, under Capability Category Ill of this                                             and any other identified requirement), by using conservative                                               failure modes in the assumptions; OR                                                                   identification process.)
b) NOTE that these mechanisms are not                                             Note that the qualitative included inthe scope of the evaluation,                                           assessment is a No discussion has been provided for the                                           requirement of the NRC impact due to the additional flood failure                                        Clarification of this SR.
mechanisms.
IFSN-A8-   This SR states: IDENTIFY inter-area              Although the obvious             Documentation of less       Appendix B includes this info       IF-IFSN-A8 Internal flooding 01         propagation through the normal flow path         propagation pathways (e.g.       obvious possible            already. Added to Section 3.1.3 -              is not from one area to another via drain lines; and    doors, stairwells, grating) were propagation pathways        "These pathways are listed in                  considered in areas connected via back flow through drain      identified, a good discussion    needs to be addressed.      Appendix B under the 'Drainage'                the fire PRA.
lines involving failed check valves, pipe and    associated with less obvious                                 section of each zone."
cable penetrations (including cable trays),      pathways (e.g. failed backflow doors, stairwells, hatchways, and HVAC           check valves, cable ducts. INCLUDE potential for structural failure  penetrations, cable trays, etc.)
(e.g., of doors or walls) due to flooding loads, for individual zones was not found.
IFSO-A1-    Based on a confirmatory walkdown performed      Appendix B contains walkdown     It is recommended that      The chilled water system operates   IF-IFSO-A1 Internal flooding by the Peer Review Team, the                    documentation which verifies     the analyst ensures that    at very low pressure and the lines             is not Revision 0                                                                                                                                                                      Page U-16

0 Florida Power & Light                                                                                                                   Aftachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                      Description                           Significance             Possible Resolution        Plant Response or Resolution          SR         Impact on Fire PRA 01          locations/impacts of some pipes containing that the plant information used   soatial information be       are insulated. Drecludino the                 considered in water may have been overlooked in the       reflects the as-built, as-       captured appropriately for   possibility of a spray. This                  the fire PRA.
analysis.                                  operated plant to support         spray concerns.              information was added to the development of flood areas.       Equipment has been           scenario description.
Based on a confirmatory plant    identified in walkdown walkdown by the Peer Review       sheets for elevation, but Team, however, it appears that    not spatial location.
some piping containing water,     Additionally the analyst and the potential impact of that  should ensure that all piping may have been              potential fluid sources in a overlooked. Some specific         given flood area are examples are provided below:      identified, and all potentially impacted In the analysis, the 4KV Room     equipment is identified was screened from further         and the impact of it failing evaluation stating that the       is evaluated.
pipes in the room do not run directly over the top of the load centers, so no spray damage is expected. However, during the confirmatory walkdown, it was identified that there is Chilled Water piping running along the North wall. This piping runs directly over the Remote Shutdown Panel in the room, and about 3 ft north of several 4160 V load centers. A failure of this line has the potential to directly impact the Remote Shutdown Panel and cause a complicated reactor trip (loss of control of various functions from the Control Room due to spurious actuations from the RSP). Additionally,, the location of the Chilled Water lines is such that a failure of the piping could result in a spray-induced failure of the 4160 V load centers due to water intrusion through the vents in in the panels, or a failure of the cooling fans on the back side of the load Page U-17 Revision 00 Revision                                                                                                                                                                   Page U-17

Florida Power & Light                                                                                                                      Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                      Description                             Significance             Possible Resolution           Plant Response or Resolution           SR        Impact on Fire PRA centers/transformers, so it is inappropriate to state that "no spray damage is expected",
and impacts from spray damage need to be identified and scenarios evaluated.
Several other electrical equipment rooms were also screened out with a comment similar to that for the 4 KV rooms. Based on the confirmatory walkdowns, Chilled water lines that had the potential to cause spray damage and resulting equipment malfunctions were also found to be present in other electrical rooms.
Fire protection piping was noted in the CCW pump room, but a discussion of its presence/impact in the room was not found in the Internal Flooding analysis.
IFSO-A3-   No process by which screening was              No formal screening seems to     PTN should document             As mentioned in IFPP-B1, no        IF-IFSO-A3 Internal flooding 01         performed is present in the analysis.          have been performed.             justification for screening     screening was performed, all areas            is not Additionally, it is difficult to particular flood areas from     were considered.                              considered in identify treatment of rooms      further analysis.                                                             the fire PRA.
without flood sources, and if this was treated appropriately.
For example, EDG42-140 is not identified to have any flood sources in the walkdown documentation. Flood propagation is evaluated for this room, but it is not discussed further in the analysis.
IFSO-A4-   No human-induced mechanisms have been          No human-induced                 It is identified that tank      As mentioned in Section 3.1.2,     IF-IFSO-A4 Internal flooding 01         included in the analysis, and additionally, no mechanisms have been             overfills will relief to vents, human-induced mechanisms are                  is not process which justifies their exclusion was    considered or identified as      drains or the waste             already taken into account in the             considered in Revision 0                                                                                                                                                                       Page U-18 40

0 Florida Power & Light                                                                                                                        Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                                Significance              Possible Resolution       Plant Response or Resolution            SR        Impact on Fire PRA provided,                                           such in the analysis.             disposal system but itis   general failure data.                          the fire PRA.
Additionally no human-induced      recommended that flooding initiators have been     specific instances be modeled as part of the internal   discussed as it relates flooding analysis. It has been    specifically to operator identified that CRs have been      induced failures.
reviewed which confirmed that     Additionally, a process or no human-induced flooding         program should be events have occurred at PTN.      identified which prevents However, no process or            human-induced floods program has been identified       from occur, thereby which would prevent such           justifying their exclusion events from occurring,            from the analysis.
IFSO-A5-   No summary or characterization of flood             No characterization of flood      Characterize flood         The flooding calculations were      IF-IFSO-A5  Internal flooding 01         sources included in the analysis has been          sources has been made (i.e.,       sources interms of         added to the notebook along with               is not provided. It is difficult to tell what the          capacity of source, pressure,     capacity, flow rate,       added discussion inSection 3.2.                considered in decisions making up the source                     temperature, etc.). Additionally   pressure, temperature,                                                    the fire PRA.
characterization were.                              no justification is given for the etc. Additionally, flow rates chosen or the           document the justification amount of time during which a      for a given flow rate. PTN flood source remains               should also document the unisolated (or what the system    process used to identify capacity is). Additionally no      potential flood sources.
calculation is provided documenting flood heights for a given scenario.
L2-1       Hot leg creep rupture was assumed to occur         This is a level C. It may not      it should be realized that In the updated Level 2 analysis     LE-F3      F&O fully 99.99% of time for all sequences, regardless       have an effect on LERF, but is     this assumption has a      (Calculation PTN-BFJR-99-019,                  addressed. No of the RCS pressure. This is based on              mentioned to indicate the          dominating effect on the   Rev. 1), the probability of hot leg            impact.
interpretation of latest NRC research which        importance of this assumption,     results. Sensitivities     creep rupture is 95%, not 99.99%, a implied that natural circulation of core debris    All other high pressure related    should be performed as     substantial difference, making a would occur at the time and with such              containment failure                needed to investigate      sensitivity analysis unwarranted.
efficiency that hot leg would fail prior to vessel phenomena are neglected            alternate assumptions.
failure and steam generator tube rupture.          because of this assumption.
Although hot leg rupture may indeed be likely, the way it is modeled precludes any other high pressure phenomena, including DCH, Induced tube rupture.
MU-2       STD-R-001 has a requirement for signoff by         Although preparation and           Consider augmenting        There is now a comments and         QU-F1      F&O fully the preparer, an independent reviewer and          review of PRA documentation        existing RRAG processes    resolutions section in the model               addressed. No the RRAG supervisor. The PRA update                takes resources away from          by defining such           update and other PRA Group                     impact.
calculations that were reviewed had all the        "productive" efforts, the          measures as a standard, Revision 0                                                                                                                                                                         Page U-19

Florida Power & Light                                                                                                                     Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                          Description                          Significance             Possible Resolution        Plant Response or Resolution           SR       Impact on Fire PRA required signatures. However there were no    elements of a PRA that is        expanded level of detail in  calculations.
review notes or discussion of the disposition being used to support plant      description of PRA           Details of each change are of review comments in the various             applications must be carefully   changes being                documented in the PTN Change calculations examined by the peer reviewers,  checked and shown to be          incorporated, items to be    Database. Each model update Further, the peer reviewers found examples    correct.                         checked for by reviewers,    includes a table of the changes of inconsistencies inseveral signed-off                                        etc. In particular, the      implemented for that model update notebooks (e.g., Accident Sequence                                             purpose/basis for each       and reasons for those changes.
Notebook included incorrect success criteria                                   change should be defined for S2 LOCAs), and examples of errors                                           in the change packages.      Consistency issues between the carried through several PRA Update Calc                                        This should provide the      Accident Sequence Analysis, the revisions (e.g., CDF cutsets that included                                     reviewer with enough         Success Criteria calculation, and single failures in emergency boration pumps                                    information to determine if  the model update calculations have for ATWS, which should have required                                            the detailed changes         been resolved in RG 1.200-related multiple failures),                                                            actually are sufficient to   upgrades of all of these documents.
fully address the basis for  The other issues are treated the change.                  separately in other F&Os.
MU-4        STD-R-002 requires a data update every 5      The RRAG is not in compliance    Update the reliability data. Data updated in current PRA model   QU-F1     F&O fully years. However, it does not appear that        with its own quality procedure   Ensure adherence to          using the data documented inthe               addressed. No Turkey Point has updated the reliability data                                   procedures                   data update calculation, Rev. 1 of            impact.
since 1995 even though the common cause                                                                      PTN-BFJR-02-026.
failure data, the human factors data, the initiating event data and the unavailability data was updated in 2000.
MU-5        STD-R-002 includes a set of criteria for when Ifthe cumulative impact of       Revise the procedure,        Changed procedure to have a         QU-F1     F&O fully to perform a model update. The key criteria    pending changes becomes          STD-R-002 to incorporate     maximum interval of five years                addressed. No are if the change has a significant impact on significant without triggering a a fixed update equivalent    between model updates.                        impact.
CDF or risk insights. However, there are no   model update. The veracity of    to the fixed data update     Whether minor changes constitute criteria as to what constitutes a significant the model is questionable for    schedule to preclude an      justification for a model update is impact. Given that there is no fixed update   any application,                 unending build up minor      determined by the model custodian.
period, there is a concern that the number of                                  changes that could have a
            'minor" changes pending could build up until                                   significant cumulative the combined impact is significant without                                      impact. The procedure triggering an update.                                                           should also be revised to include a process for evaluating the potential cumulative impact of pending changes and triggering a model update ifthe cumulative impact of the pending changes is judged to be significant Page U-20 Revision 0                                                                                                                                                                  Page U-20

Florida Power & Light                                                                                                                        Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance              Possible Resolution        Plant Response or Resolution             SR     Impact on Fire PRA QU-2        The guidance provided for the quantification     The absence of guidance          Consider establishing       Changes to the mutually exclusive      QU-F2   F&O fully process is IPE vintage. Several new codes        establishing the criteria for    clear guidance for the      event combinations, flag file, and             addressed. No are being used inthe current process which       selecting mutually exclusive     selection of mutually       recovery rule file that have occurred          impact.
did not exist at the time of the IPE. No         events, flag events, and the     exclusive event             since 1998 are documented in the guidance procedures currently exist to control   truncation level can allow the   combinations, the use and   PTN Change Database and the key processes involved in model integration      introduction of errors affecting development of flag files   model updates. They document the and quantification such as:                     the quantification results. The   to control the quantified   mutually exclusive event
: a. Criteria for development of the mutually      review of the mutually           system configuration, the   combinations, flag file, and recovery exclusive events file                            exclusive events file indicates   determination of an         rule file with justification for their
: b. Selection of truncation value                that some complimentary           appropriate truncation      content. Truncation level is set as
: c. Quantification on a sequence basis versus     combinations related to AFW      level, and acceptable       low as the hardware and software quantification of top gate                        pump maintenance might not       inputs for performance of   will allow, or until convergence is
: d. Process for breaking circular logic in the     be included. While this would    uncertainty analysis,       achieved. Uncertainty analysis input single top linked fault tree (i.e., selection of lead to conservative results                                  is described in the model update proper gate level for performing the logical     due to failure to remove invalid                              calculations.
break, naming scheme to be used for gates in     cutsets, the addition of the new system fault tree, etc.                  inappropriate mutually
: e. Process for development of flag files for the exclusive combinations would baseline quantification to ensure the            have the opposite result.
quantified configuration represents normal plant operation practices
: f. Selection of parameters for input to the UNCERT code for uncertainty calculation QU-3       The quantification of a linked fault tree model  The introduction or errors into   Consider developing a       See QU-2.                              QU-F2   F&O fully involves the proper integration of several files the quantification flag file, the documentation package                                                      addressed. No which can affect the results. For example:       mutually exclusive event file or  for the flag file, mutually                                                impact.
: a. The quantification flag file is used to set   the recovery rules file can       exclusive events file and logic flag events true or false to represent     result in elimination of cutsets  the recovery rules which normal system alignment. At PTN, this flag       from the quantification results,  provides the basis of each file is also used to set certain maintenance     The absence of valid cutsets is   item inthe respective events false,                                    typically more difficult to       files. Cross-disciplinary
: b. The mutually exclusive file is used to        recognize than the presence of    review of the flag file and remove cutsets from the results file which       invalid cutsets. Therefore,       mutually exclusive events contain certain combinations of events           thorough documentation and        file by plant personnel representing disallowed maintenance or           review of these files is          may also be considered.
illogical event combinations (i.e., events for   important to ensure that the failure to open and spurious opening of the      quantification output is same valve in a single cutset).                  providing reasonable results.
: c. The recovery rule file is used to add recovery events to the cutset results based on the appearance of certain combinations of failure events. At PTN, this process is also used to apply human error factors to the quantification results.
Revision 0                                                                                                                                                                     Page U-21

Florida Power & Light                                                                                                                 Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                          Description                          Significance           Possible Resolution       Plant Response or Resolution            SR     Impact on Fire PRA Since these files control vital processes during quantification, independent review and thorough documentation is needed to ensure that the quantification results do not exclude valid failure sequences. The current mutually exclusive events file (PTN2KMEE.TXT) was changed as a result of the addition of new T&M events for LC/SWGR HVAC AHUs and Sump Level Indicators. The calculation package includes a description of "add double maintenance events for these basic events to mutually exclusive events." However, no justification for making the events mutually exclusive or specifying the combinations that are mutually exclusive is provided. In addition, the review of the mutually exclusive events file indicates that some complimentary combinations related to AFW pump maintenance may not be included. While this would lead to conservative results due to failure to remove invalid cutsets, the addition of inappropriate mutually exclusive combinations would have the opposite result.
Similar errors can be introduced through the recovery file through the inappropriate application of recovery events to sequences which do not represent the conditions assumed inthe HRA analysis.
QU-5       Documentation was not available to indicate     For PRA applications, the NRC To meet the grade 3         In the PRA model updates,            QU-F4   F&O fully that PTN has performed qualitative evaluation   stresses an understanding of  criteria for sub-elements   sensitivity analyses are run to show         addressed. No for causes of uncertainty, such as:             sources of uncertainty in the QU-27 and QU-28             the effect of key modeling                   impact.
: a. possible optimistic or conservative success  PRA model that may be         (uncertainty analysis       assumptions. Parametric criteria,                                       important to specific         criteria), perform and      uncertainty is addressed in the
: b. suitability of the reliability data,         applications, and some        document a more             model updated calculations.
: c. possible modeling uncertainties              characterization of how the   systematic uncertainty      Comprehensive uncertainty analysis (asymmetry or other modeling limitations due    more important uncertainties  evaluation to identify      evaluations are provided in the to the method selected),                        affect the risk-informed      potential uncertainties due Uncertainty Analysis Notebook.
: d. degree of completeness in the selection of   decision.                     to such items as modeling initiating events, and                                                        assumptions, success
: e. possible spatial dependencies.                                             criteria conservatisms, data, etc, Identify the likely impacts of these sources of uncertainty on results, and perform Revision 0                                                                                                                                                              Page U-22 0

Florida Power & Light                                                                                                                       Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                              Significance             Possible Resolution        Plant Response or Resolution           SR       Impact on Fire PRA sensitivity analyses as appropriate to achieve an understanding of whether/how they may affect risk-informed decision-making using the PRA.
QU-6       Recovery of offsite power is applied to           This is a level B, because it   Verify that the recovery of  Inthe current model update, the     AS-B7     F&O fully sequences where offsite power may not             occurs throughout all           offsite power is a)         recovery of offsite power is not               addressed. No recover the lost function. This occurs intwo      sequences.                      applicable, b) will recover credited for cutsets where the                 impact.
types of circumstances:                                                           (with high probability) the recovery of offsite power does not
: 1) for non-grid-loss initiators, LOSP (and                                        lost function, and c) has   recover mitigating equipment SBO) can occur due to failure inthe AC                                            an applicable probability,  sufficient to avoid core damage.
power distribution system. XROSPi is applied.                                                                 Recovery of offsite power is not The recovery probability of XROSPi is based                                                                   credited for sequences where the on the NSAC document for restoration of                                                                       LOOP simply acts as a reactor trip offsite power to nuclear plants. The sequence                                                                 initiator (and loss of MFW) versus a inquestion is cause by a failure of a breaker                                                                 support system loss.
or transformer at the plant. It is not clear that the recovery probability is applicable.
: 2) for some SBO sequences where all SG heat removal is lost, XROSPi is applied.
Although AC power is available, it is not clear that SG Heat Removal is restored. The SG heat removal is provided by 3 TD pumps and 1 diesel driven pump. Ifthese fail, restoration of AC power does not make them operable.
For example, there is a cutset at 1.274E-9 which is loss of 4KV 3A with failure of aux transformer. There is a recovery for XROS19 and EHFPXTIE.
The correctness and reasonableness of this practice is questioned.
QU-8       The subtier criteria for a grade 3 on this        Use of PRA results for          Consider expanding the      a. A comparison of PTN CDF           QU-D4     F&O fully element considers the following to be             applications involves not just  discussion of the           cutsets to Robinson's CDF cutset               addressed. No indicative of a good understanding of the         the calculation of the core     quantification results in   was made and is documented inthe               impact.
dominant risk contributors:                       damage frequency, but an        the calculation packages    Quantification Notebook. Where
: a. The accident sequence results by               understanding of the            or developing a PSA         differences in the cutsets occurred, sequence, sequence types, and total should        significance of the dominant    Summary Document            they could be explained by design be reviewed and compared to similar plants        cutsets. Although some may      containing this type of     or data differences.
to assure reasonableness and to identify any      view this as only a             evaluation for each         b. A list of the top 50 cutsets is exceptions.                                       documentation issue, the        revision,                   provided in the model updates.
: b. A detailed description of the Top 10 to 100    process of developing this type Revision 0                                                                                                                                                                    Page U-23

Florida Power & Light                                                                                                                Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance             Possible Resolution Plant Response or Resolution               SR   Impact on Fire PRA accident cutsets (CAFTA or NUPRA) or             of documentation of the                                c. Initiating event pie charts, system accident sequences (RISKMAN) should be           quantification results often                           importance charts, and a table provided because they are be important in        leads to identification of cutsets                     listing the individual sequence ensuring that the model results are well         affecting the results which do                         contributions are included in each understood and that modeling assumption          not make physical sense.                               model update calculation.
impacts are likewise well known.                 Therefore, the practice of
: c. The dominant accident sequence groups or      documenting the quantification functional failure groups should also be         results serves a vital function in discussed. These functional failure groups       ensuring the continued quality should be based on a scheme similar to that      of the PSA model.
identified by NEI in NEI 91-04, Appendix B.
There is no discussion of results inthe calculation packages for updates provided to the review team to indicate that this type of evaluation is done of the quantification results. Also, the calculation packages provide no discussion of how the dominant cutsets or important systems were affected by the changes to the model when compared to the previous revision.
ST-1       ISLOCA Analysis:                                 This is a level A, because it                          The ISLOCA analysis has been           IE-C14 F&O fully The ISLOCA analysis was performed in 1991        may represent an ISLOCA path                           completely revised and is                     addressed. No as part of the IPE and has not been updated      that is higher than any path                           documented in calculation PTN-                impact.
since. It is currently undergoing an update. All defined to date.                                       BFJR-07-003, Rev. 1.
containment penetrations were identified for potential ISLOCA paths. Each penetration was screened or considered for further analysis.
Penetration #1 contains the RHR shutdown cooling line. There are 2 normally closed MOV's (751 and 750) in the 14 inch line. An incorrect assumption about the break location led to the screening of this penetration. This assumption would not have occurred if the structural strength of the entire pipe segment was considered.
The RHR pipe is designed for 2500 Ib,to the discharge of MOV 751. The pipe is then 600 lb pipe to the RHR suction. The pipe is then 150 lb pipe back to the RWST. There are no check valves in the line from the RHR MOV 751 to the RWST. Ifthe tenets of NUREG/CR-5744 are followed and the ultimate strength of the pipe is considered, Revision 0                                                                                                                                                             Page U-24

Florida Power & Light                                                                                                                   Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                         Description                           Significance             Possible Resolution       Plant Response or Resolution           SR     Impact on Fire PRA the pipe rupture location will not likely be inside containment. Ifa realistic analysis is performed, a portion of these events will not result in any pipe rupture at all, causing a back flow from the RCS to the RWST.
SY-2       The RHR/LHSI model for recirculation            RWST level is the primary cue    The discrepancy between    Added JHFA3RWSTLVL, "common         HR-C3    F&O fully assumes that failure of either the RWST level  for switchover to recirculation. definition of dependencies cause miscalibration of U3 RWST              addressed. No indication or the sump level indication will   A number of other plant PRAs     in section 3.6.5 and the   level indicators" to gate J504.              impact.
result in a failure to switchover to           indicate common cause            model should be resolved   Added JHFA4RWSTLVL, "common recirculation. The model for failure of the    miscalibration of RWST level     and the model revised      cause miscalibration of U4 RWST sump level indication includes common cause     instrumentation to be a          accordingly.               level indicators" to gate U4J504.
miscalibration but common cause                reasonably important miscalibration of the RWST level indicators is contribution to CDF. The not included in the model and no basis could   Robinson Nuclear Plant CCW be found for the exclusion of this failure. system recently experienced a The CCW model does not include the relief       relief valve opening coincident valve, or the surge tank level instrumentation, with maintenance on the surge tank level indication. This is a precursor to loss of CCW. This event should be reviewed for application to Turkey Point.
TH-1       The basis for requiring or dismissing HVAC      A consistent approach should     Consider revising analysis The HVAC analysis has been fully    SY-B6    F&O fully requirements is poorly supported and            be developed for HVAC            to better support HVAC     revised. GOTHIC room heat-up                 addressed. No inconsistent throughout the PRA. The            requirements for AC,DC and       requirements.              calculations have been performed             impact.
following observations were made:               EDG rooms. This should be                                   for many of the HVAC-cooled areas
: 1) The HVAC system notebook and DG              supported with applicable                                   containing PRA components, system notebook require HVAC for EDG            analysis. Now that it is known                              especially those where the need for rooms, DC equipment room and 4160               that unit 3 DG HVAC                                         HVAC was questionable. In some equipment rooms. The analysis is based on       requirements are different from                             cases, the heatup calculation design basis calculations from A&E done in      the unit 4 DG HVAC                                          results were compared to the 1985-1988.                                      requirements, it may not be                                 survivability temperatures of the
: 2) Recent updates use engineering judgment      appropriate to use the same                                 PRA components inthe various and plant experience from system engineer to    model for both units.                                       rooms and the dependencies dismiss need for HVAC to 4160 and DC                                                                        applied accordingly, power rooms.
: 3) The current fault trees require HVAC for DC room and EDG room for LOCA events only.
: 4) GOTHIC or other room heat up calculations have not been done to support room cooling of these rooms.
Discussions with plant staff during the Revision 0                                                                                                                                                                Page U-25

Florida Power & Light                                                                                                                      Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                          Description                             Significance             Possible Resolution        Plant Response or Resolution            SR   Impact on Fire PRA Certification indicate the following requirements for HVAC:
A) Unit 3 DG does not need HVAC.
B) Unit 4 DG need HVAC whenever they operate C) the switchgear room needs ventilation to protect the 480v transformer. The switchgear ventilation system is normally running.
Remedial action via opening doors and running an exhaust fan is sufficient to maintain temperatures. The lead time and indication of loss of ventilation is sufficient enough that loss of switchgear room ventilation is not considered an initiating event.
D) The inverter and battery charger need room cooling. Remedial action is available with plug-in, portable fans, but recovery time is on the order of 1 hour.
TH-3       It is difficult to determine, from the PSA
* Clearer guidance should be     Provide clearer              The updated Accident Sequence       SC-B1  F&O fully notebooks, which codes or methods of                 provided to preclude potential traceability of success      Analysis Notebook (Revision 3), the        addressed. No analysis are used for specific success criteria      miss-application of T/H codes  criteria to analytical       new Success Criteria Calculation           impact.
determination, or why these methods are              and assumptions.               bases, at least for "non-    (PTN-BFJR-08-014, Rev. 0), and appropriate. For example, applications of the        The analytical bases for the   obvious" criteria (e.g., for the model update calculation (PTN-MAAP code, particularly the IPE-vintage 3b           success criteria should be     those whose basis is         BFJR-00-001, Rev, 9) resolved this version, may require some justification or           clear.                         other than FSAR).            F&O.
check for applicability (e.g., avoiding use of                                      Consider developing a MAAP 3.Ob for rapid RCS depressurization                                            clear set of guidelines scenarios, which typically require capabilities                                     establishing the beyond what was available in that particular                                        acceptable range of version of the code).                                                               applications of various Further, it is difficult to determine the specific                                  types of codes and analytical bases for specific success criteria                                      calculations.
used in the model. While the Accident Sequence Notebook includes a summary of success criteria for each event, reference for the bases for the success criteria is to the IPE, which does not provide additional information on this subject.
TH-4       The LOCA break size definitions for the PRA          The definition of LOCA break   Because the break size       The updated Initiating Events       IE-A2  F&O fully are based on different criteria than those for       sizes has become somewhat      definitions are central to   Notebook (Revision 3), the updated         addressed. No most other PRAs. This is acceptable ifthe            standardized across plant      the LOCA modeling for        Accident Sequence Analysis                 impact.
underlying analyses provide sufficient basis         PRAs since the IPEs. There is  the Turkey Point PRA,        Notebook (Revision 3), the new Revision 0                                                                                                                                                                   Page U-26 0

Florida Power & Light                                                                                                                       Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                             Significance               Possible Resolution        Plant Response or Resolution         SR        Impact on Fire PRA for the definitions.                            no requirement that size range     there should be, in the      Success Criteria Calculation (PTN-A series of MAAP 3.Ob analyses was              definitions follow any "standard   event tree notebook or       BFJR-08-014, Rev. 0), and the performed for the Turkey Point IPE. The        convention," but ifthey do not,    appendix, a clear            updated Revision 8 PTN PRA available documentation consists of the        it is important that they be       discussion of the bases      model (PTN-BFJR-00-001, Rev. 9)
Accident Sequence Notebook descriptions         carefully defined and have a      for the selections,           provide justification for the LOCA and success criteria summary and an internal    sound basis that is supported      including reference to the   break sizes.
memo from the IPE, which provides a            by realistic analyses. Some       spectrum of analyses summary listing of the MAAP cases that were     analyses were performed for        performed and the run, along with an indication as to whether or  the Turkey Point IPE success      specific set of MAAP or not core uncovery/vessel failure occurred.      criteria. However, the analytical  other analyses that define Reviewer note R11 to table TH provides a       basis for a key element of the    the size range for each comparison of the definitions and their bases,  above definitions was             size break.
with focus on the injection phase, as           unavailable. While the            Consideration should be discerned from this information:                reviewers believe this is an      given to evaluating and From the comparison in note R1 1, it can be     important observation, it affects documenting the effect on seen that the principal difference in size      mainly the larger LOCAs,           PSA results and risk definitions (aside from the names used) is in   which are not dominant            insights resulting from the PTN Medium Break category, which is         contributors to the CDF and       using these (as opposed essentially the lower end of the typical Large  LERF results, and so a            to more "traditional")
Break category.                                 significance B is appropriate. definitions.
Comments on the above are as follows:                                             Confirm that all definitions The available documentation provides the                                          are based on analyses basis for some, but not all, of the size ranges                                   performed using noted above. Information provided in FPL                                          appropriate codes and memo NF-90-450 (October 19, 1990)                                                 modeling assumptions, provides sufficient information to serve as a                                     especially for the larger basis for the S1 and S2 ranges and the lower                                      break size definitions (i.e.,
end of the Medium LOCA range. But it does                                         those in the range of 3" not provide any basis for the upper end of the                                    and above). Consider Medium LOCA / lower end of the Large LOCA                                         confirming the results of size ranges (i.e., the 13.5" break). Available                                    key earlier MAAP 3.Ob MAAP runs listed in the memo are for breaks                                       analyses against results up to 10" diameter. Discussions with FPL                                          obtained using currently personnel identified that the 13.5" size cutoff                                   available versions of may have been selected by the IPE                                                 MAAP, which have contractor during the early stages of the IPE,                                    improved capabilities for but a specific basis was not located during                                       modeling depressurization the review.                                                                       and other T/H For the TPN Medium LOCA, i.e., breaks up to                                       phenomena.
13.5 inches, the PRA assumes that a single train of high head injection can mitigate this class of LOCAs, whereas typical PRAs would instead tend to credit a single train of low head injection for breaks at the upper end of this size range (i.e., above 6"). As noted Page U-27 Revision 0                                                                                                                                                                    Page U-27

Florida Power & Light                                                                                                Attachment U - Internal Events PRA Quality Table U Turkey Point PRA Facts and Observations F&O                        Description                    Significance           Possible Resolution   Plant Response or Resolution    SR        Impact on Fire PRA above, analyses supporting the upper end of the Medium LOCA range with this success criterion were not available during the peer review.
MAAP 3.Ob analyses were used to support the definition of ECCS requirements for the MLOCA, even at the upper end of the break size range (i.e., 13 inches). Ingeneral, MAAP 3.Ob is not appropriate for rapid depressurizations as would be occurring for breaks in the MLOCA size range.
Page U-28 Revision 0 Revision 0                                                                                                                                             Page U-28

Florida  Power & Light                      Affachment tahetV-Fr V - Fire PRA R
Clualitv ult Flrd oe      ih V. Fire PRA Quality 82 Pages Attached Revision 0                                                        Page V-1

Florida Power & Light                                        Attachment V - Fire PRA Quality V.1     Fire PRA Quality Overview                                                            0 The Turkey Point (PTN) Fire PRA Peer Review was performed in February 2010 at PTN using the NEI 07-12 Fire PRA peer review process, the combined PRA standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2. The purpose of this review was to provide a method for establishing the technical quality and adequacy of the Fire PRA for the spectrum of potential risk-informed plant licensing applications for which the Fire PRA may be used. The February 2010 PTN Fire PRA Peer Review was a full-scope review of all the Technical Elements of Section 4 of the ASME/ANS standard. This report was issued to PTN in April 2012. A subsequent peer review performed in March 2012 was a focused scope peer review addressing the FSS, HRA and PRM Technical Elements. The report was finalized and issued to PTN in May 2012.
The Fire PRA update addressed the Supporting Requirement assessed deficiencies (i.e., Not Met or CCI). Completion of recommendations related to Supporting Requirement assessments and 'Finding' F&Os results in a Capability Category II assessment for the majority of the Supporting Requirements. The basis for acceptability of Capability Category I requirements is documented in this attachment.
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.
Based on the completion of peer review recommendations and the assessment of deferred items, the PTN Fire PRA is adequate to support the NFPA 805 Fire Risk Evaluation process.
Table V-1 provides an assessment of the PTN Fire PRA quality by supporting requirement.
Table V-2 provides the supporting requirements that are assessed as 'Not Met' or 'CC I' and discusses the potential impact of each on the application.
Table V-3 provides the peer review findings and the disposition for each in the Fire PRA.
V.2     Electrical Panel Factor Methods The fire PRA used a method for defining factors for electrical panel fires that was considered as an unreviewed analysis method at the time of the last focused scope peer review (F&O 10-3 for SR FSS-C4, FSS-D5, and FSS-G1). Since that time the method has been accepted by the EPRI panel. However, the NRC member of that panel dissented on the panel's approval of the method. Therefore the method is described below and a sensitivity analysis was performed comparing against a NUREG/CR 6850 method. For the areas with the highest risk using this method additional defense in depth has been specified.
The table below provides severity factors that can be applied to different types of ventilated/open cabinets for the case in which targets within the zone of influence would be damaged. A separate non-severe scenario is also defined for these panels in which the panel itself is damaged and any components with cables terminating at the panel are also damaged (including potential spurious operations).
Revision 0                                                                          Page V-2

Florida Power & Light                                            Affachment V - Fire PRA Quality Panel Type                  Conditional Probability of Propagating Fire Switchgears                                   2.5E-01 Load Centers                                  2.5E-01 Motor Control Centers                              2.OE-1 Low Voltage Panels                               1.OE-01 The severity factors used for electrical cabinet fires were derived in the Supplemental Fire PRA Methods report which was submitted to the EPRI Fire PRA Methods Review Panel. This review was not complete as of the date of the completion of the fire PRA.
The anticipated conditional probabilities of a propagating fire for cabinets included in Bin 15 which was expected to be endorsed by this panel are shown in the above table.
When developing scenarios, the 'recommended' conditional probabilities were applied to open severe fire scenarios defined as fires impacting targets outside of the ignition source and a conditional probability of 1.0 was applied to non-severe, NS, fires impacting only those cables terminating within the ignition source.
A manual non-suppression probability associated with the time required to form a hot gas layer is used in conjunction with the above factors to define the probability of hot gas layer formation. In order to ensure that the manual non-suppression probability used, which is derived from the fire events database, does not credit suppression effects associated with the derivation of the above panel factors, a reduced level of credit for the manual non-suppression probability is used. The nominal mean suppression rate used to derive the manual non-suppression probability in FAQ 08-050 for electrical panel fires is 0.102 min-'. A value of approximately 10% below the mean suppression rate specified for all fires (in Table 2 of FAQ 08-050), 0.06 min-', was used in conjunction with the above factors to compensate for any potential double counting of the non-suppression effects between the panel factor development and the manual non-suppression probability. This mean suppression rate is conservative given that it is representative of that specified for oil fires which are significantly more aggressive than the panel fires and thus provides a significant level of conservatism in order to ensure that any double counting in the non-suppression probability is eliminated.
The results of the sensitivity analysis, replacing the above panel factors with a conservative factor of 1.0, reflect an increase in the probability of HGL scenarios and associated zone of influence and multi-compartment analysis impact. The impact on the fire zone baseline risk quantification (for CDF and LERF) was found to range from an increase of 15% to 35%. This increase along with the impact on HGL probability and delta CDF/LERF is considered conservative as it completely eliminates the panel factor.
The actual results are expected to be between the current quantification and the values obtained with a panel factor of 1.0.
Page V-3 Revision 00                                                                                Page V-3

Florida Power & Light                                                    Attachment V - Fire PRA Quality Note - Grey shading in the below identifies those elements which were the subject of the March 2012 focused scope peer review.
Table V-1 PTN Fire PRA Quality Summary Supporting Requirement                    Peer Review Capability Assessment PP-Al                                        Cat 1-3 is MET PP-C2                                         Cat 1-3 is MET PP-C4                                         Cat 1-3 is MET ES-A5                      Cat 1 is MET; Cat 2 is MET; Cat 3 is Not Assessed ES-A6*                     Cat 1 is MET; Cat 2 is MET; Cat 3 is Not Assessed ES-B1                      Cat 1 is MET; Cat 2 is MET; Cat 3 is Not Assessed ES-B2                      Cat 1 is MET; Cat 2 is MET; Cat 3 is Not Assessed ES-B4                                         Cat 1-3 is MET CS-Al                                        Cat 1-3 is MET CS-A2                           Cat 1 is MET; Cat 2 is MET; Cat 3 is MET CS-A3                                         Cat 1-3 is MET CS-A4                                         Cat 1-3 is MET CS-A5                                         Cat 1-3 is MET CS-A6                                        Cat 1-3 is MET CS-A7                                         Cat 1-3 is MET CS-A10                      Cat 1 is MET; Cat 2 is MET; Cat 3 is Not Assessed CS-Al1                                         Cat 1-3 is MET CS-R1                                 Cat 1 is MET; Cat 2-3 is MET CS-cl                                         Cat 1-3 is MET CS-C2                                         Cat 1-3 is MET CS-C3                                         Cat 1-3 is MET CS-C4                                         Cat 1-3 is MET QLS-Al                                         Cat 1-3 is MET QLS-A2                                         Cat 1-3 is MET QLS-A3                                         Cat 1-3 is MET QLS-A4                                              N/A QLS-B1                                         Cat 1-3 is MET Revision 0                                                                                      Page V-4

Florida Power & Light                                                     Attachment V - Fire PRA Quality Table V-1 PTN Fire PRA Quality Summary Supporting Requirement                    Peer Review Capability Assessment QLS-B2                                       Cat 1-3 is MET QLS-B3                                            N/A Aat C aM MET ReiPo          MR0                                                                          Page V-5LT Aa`
Revision 0                                                                                       Page V-5

Florida Power & Light                                               Attachment V - Fire PRA Quality Table V-1 PTN Fire PRA Quality Summary Supporting Requirement                    Peer Review Capability Assessment
                           ~momm Reison0Pae                                                                                   -
SIM Revision 0                                                                                 Page V-6

Florida Power & Light                                                      Aftachment V - Fire PRA Quality Table V-1 PTN Fire PRA Quality Summary Supporting Requirement                    Peer Review Capability Assessment IGN-A1                                       Cat 1-3 is MET IGN-A7                                       Cat 1-3 is MET IGN-A8                           Cat 1-2 is MET; Cat 3 is Not Assessed IGN-A9                                       Cat 1-3 is MET IGN-B1                                       Cat 1-3 is MET IGN-B2                                       Cat 1-3 is MET IGN-B3                                       Cat 1-3 is MET IGN-B5                                       Cat 1-3 is MET QNS-B1                                             N/A QNS-D2                                             N/A evision 0a                                                                                   Page V-7ME Revision 0                                                                                        Page V-7

Florida Power & Light                                                  Attachment V - Fire PRA Quality Table V-1 PTN Fire PRA Quality Summary Supporting Requirement                    Peer Review Capability Assessment SF-A4                                       Cat 1-3 is MET SF-A5                                       Cat 1-3 is MET SF-B1                                       Cat 1-3 is MET FQ-A1                                       Cat 1-3 is MET FQ-A3                                       Cat 1-3 is MET FQ-A4                                       Cat 1-3 is MET FQ-B1                                       Cat 1-3 is MET FQ-D1                                       Cat 1-3 is MET FQ-F2                                            N/A MU-Al                                       Cat 1-3 is MET MU-A2                                       Cat 1-3 is MET MU-B1                                       Cat 1-3 is MET MU-B2                                       Cat 1-3 is MET MU-B3                                       Cat 1-3 is MET MU-B4                                       Cat 1-3 is MET MU-Cl                                       Cat 1-3 is MET MU-D1                                       Cat 1-3 is MET MU-E1                                       Cat 1-3 is MET Page V-8 Revision 00 Revision                                                                                     Page V-8

Florida Power & Liqht                                                     Attachment V - Fire PRA Quality Note - Grey shading in the table below identifies those elements which were the subject of a subsequent focused scope peer review. This table will be updated to reflect that peer review.
Table V-2 PTN Supporting Requirements Not Met or Capability Category I Supporting            PTN Final Capability                               Comment Requirement                Assessment PP-B1                  Cat 1-3 is NOT Met           The use of fire zone boundaries as PAU boundaries was of concern for the peer review team. However, the use of these boundaries was based on walkdowns of fire scenarios and inclusion of targets in adjacent zones where those targets are within a fire's zone of influence.
PP-B2                  Cat 1 is NOT Met; Cat 2-3 is See discussion under SR PP-B1 above NOT Met PP-B3                  Cat 1 is NOT Met; Cat 2-3 is See discussion under SR PP-B1 above NOT Met PP-B5                  Cat 1 is NOT Met; Cat 2-3 is See discussion under SR PP-B1 above NOT Met PP-B7                  Cat 1-3 is NOT Met           See discussion under SR PP-B1 above ES-Al                  Cat 1-3 is NOT Met           To address concerns identified by the peer review team with respect to the appropriateness of applying a reactor trip initiator for all fires a review of the mapping of components to individual component failure modes was performed to ensure that these failures were appropriately propagated in the fault tree for a reactor trip initiator ES-A3                  Cat 1-3 is NOT Met           See discussion under SR ES-Al ES-A4                  Cat 1-2 is NOT Met; Cat 3 is See discussion under SR ES-Al Not Assessed ES-Cl                  Cat 1-3 is NOT Met           Credit is taken for SSD instrumentation only as cues for HFEs. Guidance to operators to focus on these instruments during post fire shutdown supports this approach and eliminates the potential complexity of the shutdown that would result from attempting to credit all possible instrumentation cues.
ES-C2                  Cat 1 is NOT Met; Cat 2 is   See discussion under SR ES-Cl NOT Met; Cat 3 is Not Assessed Reqsio 0rl                                                                              Page V-Revision 0                                                                                                Page V-9

Florida  Power & Light Flrd oe       ih Attachment tahetV-Fr V - Fire PRA R Quality ult Table V-2 PTN Supporting Requirements Not Met or Capability Category I Supporting             PTN Final Capability                             Comment Requirement                Assessment IGN-A4                 Cat 1 is MET; Cat 2 is NOT   Fire events identified were found to be non-Met; Cat 3 is Not Assessed   challenging. Therefore, no outlier experience was found. Therefore, Cat II is considered to have been met. Peer review conservatively considered the lack of detailed information with respect to certain fire events to be a basis for considering these events challenging.
IGN-A10                Cat 1 is MET; Cat 2 is NOT   A parametric uncertainty evaluation was generated Met; Cat 3 is Not Assessed   subsequent to the peer review incorporating the IGN parametric uncertainty thereby meeting the requirements of Cat 2 for this SR.
CF-Al                  Cat 1 is NOT Met; Cat 2-3 is The peer review considered the application of circuit NOT Met                      failure probabilities to some cable failures to be inadequate. However, given that the circuit failure probabilities were applied to risk-significant contributors, this SR is considered to be MET at Cat Il/111 CF-A2                  Cat 1-3 is NOT Met           A parametric uncertainty evaluation was generated subsequent to the peer review incorporating the CF parametric uncertainty thereby meeting the requirements of this SR.
CF-B1                  Cat 1-3 is NOT Met           A more detailed basis for the circuit failure probabilities used was provided in the analysis to address this issue, therefore, this SR is now considered to be met at Cat 1-3.
ISO=                                -
nMMIMMEE Page V-1O RevisIon 00 Revision                                                                                             Page V-10

Florida Power & Light                                                  Attachment V - Fire PRA Quality Table V-2 PTN Supporting Requirements Not Met or Capability Category I Supporting            PTN Final Capability                            Comment Requirement                Assessment SF-Al                  Cat 1-3 is NOT Met          Credit is taken for IPEEE and other seismic assessments as a basis for meeting the seismic fire requirements of the standard. Seismic fire analyses are qualitative in nature and will not impact overall fire risk.
FQ-A2                  Cat 1-3 is NOT Met          See discussion under SR ES-Cl FQ-Cl                  Cat 1-3 is NOT Met          The lack of dependency evaluation for HFEs represented by altered basic events FQ-E1                  Cat 1-3 is NOT Met           Uncertainty analyses not available at the time of the peer review have been performed.
FQ-F1                  Cat 1-3 is NOT Met           Uncertainty analyses not available at the time of the peer review have been performed. Importance results have been incorporated into the summary report.
UNC-Al                 Cat 1-3 is NOT Met          See discussion under SR FQ-F1 UNC-A2                 Cat 1-3 is NOT Met          See discussion under SR FQ-F1 Page V-lI Revision 00 RevIsion                                                                                           Page V-11I

Florida Power & Liaht                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR              Basis and Recommendation                   Disposition in Fire PRA Update 1-1            In numerous significant scenarios, the          CF-Al           The overall Fire PRA results appear to be     This F&O has been resolved.
spurious operation probability is assumed to                    greatly impacted by setting spurious          At the time of the Peer Review, the FPRA be 1.0 (true) for any events where spurious                     operation probabilities to 1.0. Scenario      had only a very limited credit for fire induced operation can occur. For example, in                            79ALA, which is presently 8E-06 would be      spurious actuation probability. The specific scenario 79ALA (one of the top 5 scenarios                      reduced by at least an order of magnitude     instance identified in the F&O was updated.
in unit 3), three events are set to true                        by assigning spurious operation probabilities The resolution of this F&O also included a affecting the top cutsets; GMMOGE100                            to several events. Similarly, with 79AKA, and review of significant fire initiating events and (MOVs 878A or B spuriously operate),                            79AJA also at 8E-06.                          additional credit for hot short induced MAVK3CV303A, OHTX3CNTRL. Capability                             Perform Circuit Failure Probability Analysis  spurious operation was applied in the Category I requires setting spurious                            for significant spurious operations events,   analysis as appropriate. In all cases, the operation probabilities to industry accepted                    and modify the FRANC model to assign a        application of the spurious actuation factor is values. It appears most of the events set to                    Perform Circuit Failure Probability Analysis  consistent with the guidance in NUREG/CR-true would be either MOVs (0.33) or AOVs                        for significant spurious operations events,   6850 and FAQ 08-0047.
(0.62) or similar, and should not be set to                     and modify the FRANC model to assign a true for significant fire scenarios. Analysis                   probability for the event in the cutsets. In using the specific circuit configuration                        order to meet CCII, the spurious operation for each significant spurious operation would                   probability should be based on the specific be required for CC II, and may lead to                          circuit configuration for each significant different results than the generic values,                      spurious operation.
depending on the circuit design and cable affected.
1-10           Transient Fires are postulated in all fire      IGN-A9          Systematic issue. Appears as if numerous      This F&O has been resolved.
compartments, as listed in Appendix B and                       compartment transient frequencies were        A sensitivity evaluation was performed that Table 3-6 of the Ignition Frequency Report.                     underestimated, while others would have       involved increasing the weighting factor for All factors affecting the fire frequency were                    been slightly over estimated as a result,     occupancy and storage from 'low' to assessed based upon a slightly modified                         Initial review was confirmed by walkdown of   'medium' for all instances where such a NUREG/CR-6850 approach. However, the                            5 areas. The ranking on all 5 areas did not   condition could reasonably be expected to rankings that were provided do not appear                       appear to match the walkdown teams            occur. The results of this sensitivity found to be consistent with the methods in                            estimate for each area.                       that the impact on the calculated CDF for NUREG/CR-6850, result in an                                     Re-assess the transient fire rankings per the each unit was less than 1E-7. Given this underestimate for fire frequencies in some                      Guidance in NUREG/CR-6850. Confirm the        small impact, the existing analysis is areas, and an over estimate in other areas.                     rankings by walkdown of each area, taking     adequate for the application.
One F&O is provided on this SR. In                              into account the actual condition.
particular: a) Areas were ranked as zero in maintenance, occupancy, or storage even though entrance to the areas is physically possible, b) Areas were ranked as 1, even though activities were not prohibited by plant procedure.
Revision 0                                                                                                                                                           Page V-12 0

0                                                                         0 Florida Power & Light                                                                                                          Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                   Fire PRA SR            Basis and Recommendation               Disposition in Fire PRA Update In areas where the room is sealed during operation (roof plugs), transients could have been left inthe room prior to sealing, so the ranking on this factor should not be zero -
per the 6850 guidance. During the walkdown, Compartments 70 and 71 both had permanently stored breaker grounding devices, with poly-covers, and 71 had a temporary transformer for the polar crane (operating). Both should be ranked as
               'medium' for storage. Similarly, the cable room had storage of 3 temporary fans, cables and blankets and should be marked as medium for storage. This room also appears to include numerous components that will likely be worked on during power, (ranking moderate for non-hot work), and numerous people were present during our limited walkdown. Compartment 88, an open area infront of the switchgear room, had numerous combustibles stored and located, and should probably be marked as medium or high (presently marked as low). Both area 85 and 88 have frequent foot traffic, and should be marked as medium for occupancy. 85 appears as if it should be moderate for storage (no controls). Similarly; no controls appear to be in place for 116.
The above are samples of identified issues, based on our limited walkdown. Itappears there will be similar issues with other areas inthe plant. We looked at other areas adjacent to the areas we were in (compartments 87, 84, etc), and expect similar problems with the present rankings.
(This F&O originated from SR IGN-A9) 1-11          Events in the altered events table are         CS-Al1          Affects multiple recoveries in numerous This F&O has been closed.
recovered with a 0.1 failure probability       CS-C3           compartments                            The action taken to address this item was (generally) that include opening of valves,                    Perform 92-18 reviews on all recovered  specifically included inthe focused-scope Revision 0                                                                                                                                                Page V-13

Florida Power & Light                                                                                                                   Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                        Fire PRA SR            Basis and Recommendation                  Disposition in Fire PRA Update etc. However, these components have not            HRA-D2         components in the Altered Events Report. Peer Review.
been confirmed as available and are                PRM-B9         Consider also that some recovery events unaffected by spurious operation failing the                      may involve the manual operation of more valve (92-18 concern),                                            than one component.
(This F&O originated from SR CS-Al1) 1-17           Table 3-2 includes uncertainty values (EF)         IGN-A10        Systematic Issue.                            This F&O has been resolved.
for prior and posterior values. However,           QU-E3          Estimate EFs for significant fire            The quantitative uncertainty analysis was Error Factors are not propagated to the            UNC-A1         compartments. ESTIMATE the uncertainty       prepared subsequent to the peer review.
compartment specific ignition frequencies.                        interval of the CDF results. ESTIMATE the    A parametric uncertainty evaluation that The other parameters, such as conditional          UNC-A2         uncertainty intervals associated with        considers fire ignition frequency as well as failure probabilities for circuit failures, do not                parameter uncertainties (DA-D3, HR-D6,       other variables was performed that uses a have uncertainty intervals.                                       HR-G8, IE-C15), taking into account the      Monte Carlo sampling process. The results The lack of uncertainty intervals would not                       state-of-knowledge correlation.              of the analysis showed a mean that was generate meaningful uncertainty interval of                       s       f       e           lote                    ais       sha        meancthat was the CDF/LERF results. (This F&O originated                                                                     slighter higher than the calculated results from SR IGN-A10)                                                                                               which was expected.
1-18           During walkdowns, several key areas                IGN-A7         Appears to be missing components in          This F&O has been resolved.
appeared to have ignition sources not                             numerous areas, based on a limited           The specific instances identified in the F&O included on the ISDS. For example, in the                         sampling during walkdown.                    were reviewed and the analysis updated cable spreading room, 2 transformers were                         Perform a re-verification of the ISDS for    accordingly. In addition, the supplemental in the compartment (3X033 - 75KVA, 3X130                          significant fire areas in the FPRA. Add      walkdowns that were performed as part of
                - 45KVA), both within the screening distance                      missing components to each ISDS, where       ongoing analysis refinements efforts for the of targets. Also in the compartment is CP-                        applicable,                                  significant fire areas did not identify any 600 spectralink cabinet, an open cabinet,                                                                      other omissions.
the RCP Vibration Monitoring Cabinet, 4P21 and 4P09 instrument AC panel. Note; we did not do a 100% review of the CS room, so additional cabinets may be missing. See also F&O 1-19. (This F&O originated from SR IGN-A7) 1-19           It appears the Ignition Source Counting did        IGN-A7         Appears to be a systematic issue in the      This F&O has been resolved.
not count Lighting Panels or other similar                        FPRA.                                        A re-assessment of the lighting panels was panels. For example, there were at least 8                        Include unsealed lighting panels and similar performed. The re-assessment focused on lighting panels in the cable spreading room                       electrical cabinets in the ISDS as potential the need for treatment as a fire initiating that were not on the ISDS. Additional similar                     ignition sources.                            event. No effort was undertaken to alter the panels are located in most electrical rooms                                                                    population of electrical cabinets considered we walked down, such as the switchgear                                                                         in the fire frequency development.
Revision 0                                                                                                                                                            Page V-14

0 Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR             Basis and Recommendation                   Disposition in Fire PRA Update rooms and other electrical rooms. Based on                                                                    Therefore, the existing values potentially our walkdowns, many of the lighting panels                                                                     have a conservative bias. The assessment should be included in the ISDS, based on                                                                      did not identify any instances were explicit guidance in 6850 and the subsequent FAQ.                                                                      treatment as a fire initiating event was on sealed cabinets. A review of the generic                                                                   needed.
guidance provided for ignition counting did list the screening of small, wall mounted cabinets (sealed). However, the lighting panels do not appear to meet the criteria listed in the procedure (not sealed, numerous switches/breakers), etc. Many of the cabinets are located close to cable trays or other intervening combustibles, so a small fire could result in a larger fire due to spreading. (This F&O originated from SR IGN-A7) 1-2           Section 4.1of the Component Selection            AS-B1           The significance of not identifying           This F&O has been resolved.
Report mentions: "Since the FPRA                 ES-Al           components as causing initiating events is    The FPRA assumes each postulated fire quantification calculates a fire CCDP and the                    basically that the assumed model impact is    results in at least a reactor trip. Logic is initiating event frequency for each zone is      ESA3            accurate by modeling a reactor trip with a    included in the model so that appropriate based on the fire ignition frequency, the        ES-A4           subsequent failure of the function, rather    event tree is quantified if the fire induces a initiating event faults are not required to be   FQ-A2           than modeling the initiating event itself. In different type of event (event tree). The used for FPRA quantification." Fault tree                        some cases, this impact is a matter of timing overall structure of the FPRA model was initiating events were not impacted by the                       for operator actions. In the case of this     reviewed to address the specific item component mapping, and are therefore not                         FPRA, the HEPs have been conservatively       identified in the F&O and to confirm changed by fire damage. As a result,                             set assuming a loss of MFW as a starting      appropriateness of overall treatment. The equipment associated with Fault Tree                             point. However, the fault tree initiating     only change that was required was related initiating events were not identified as                         events include loss of CCW, loss of HVAC      to biasing the application of recovery actions components potentially causing a fire-                           and others. It is not clear that the present  so that they were based on an assumed loss induced initiating event. 163 events are                         model accurately determines CDF/LERF          of MFW.
screened in Table A of the Equipment                             results for systems impacted which may Selection Analysis based on being                                cause a complicated reactor trip (special associated a                                                     initiating event).
fault tree initiating event. Most are modeled                    Modify FPRA to model the fire impact to in other system models. However, Several                         Fault Tree Initiating Events, and analyze the were found to not be modeled in the rest of                      FPRA assuming a fault tree initiating event the model: CPSD3PC61 1, CPSD4PC611                               for those areas where the initiating event and 2 related failures. A few others (Cooling                    can occur.
units) do not appear to be modeled Revision 0                                                                                                                                                           Page V-15

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR           Basis and Recommendation                        Disposition in Fire PRA Update elsewhere.
(This F&O originated from SR ES-Al) 1-22           Events DACF3ECCB & DACF3ECCA are                HRA-Cl          A review of the Altered Events Report            This F&O has been closed.
anded under the containment cooler Fault                        indicates there are likely numerous              The action taken to address this item was Tree Logic. Each are applied with a                             combinations of scoping HEPs (0.1 in the         specifically included in the focused-scope screening value recover of 0.1, resulting in a                  altered events report.                           Peer Review.
combined recovery of 0.01 Similarly,                            Review the altered events report for AAVK3-2831 and 2832 combine under an                            significant fire areas, and determine the AND gate, resulting in a combined recovery                       combined HEPs where they occur under an of 0.01. The above are only examples,                           AND gate. Use a single recovery for these based on a random review of two fire areas,                     events, or set only one of the events to 0.1 and a few HEPs for each. Two events in                          in the altered events report.
about 15 were found to have the above problem. (This F&O originated from SR HRA-Cl) 1-24           Event EREE3286G3F is listed as 0.0              HRA-Cl          Multiple discrepancies in the altered events     This F&O has been closed.
probability in the altered events report for 2  PRM-B11         table. Some may be documentation issues          The action taken to address this item was areas, even though the description says set                     (incorrect description). However, many           specifically included in the focused-scope to .1. A query of the Altered Events Table                      appear to be errors in the probability for       Peer Review.
comes up with 146 entries, where the                            these events. Setting the events to 0.0 probability is set to 0.0, but the description                  results in the events being screened from says to set the event to 0.1 A question was                     the PRA results. However, many may be as asked on this issue, and the response was                       a result of setting other related events to 0.1.
that these were set to 0.0 and another event                    In this case, the documentation associated was set to 0.1. However, there is no way to                     with the reason/description needs to be verify, track or repeat these settings. In                      updated.
reviewing a few events, there was no easy                       Correct Altered Events Table, where way to determine that there is a                                applicable. Additionally, when events are set corresponding 0.1 event that recovers                           to zero, provide a traceable method to the failure. (This F&O originated from SR                       determine that a corresponding event is set HRA-Cl)                                                         to 0.1 or another value.
1-25           There does not appear to be a review of         FQ-E1           Requirement of QU-D5 as called for by FQ-        This F&O has been resolved.
non-significant cutsets in the PRA              QU-D5           El                                               Review of non-significant cutsets performed documentation.                                                  Perform a review of non-significant cutsets      and documented.
(This F&O originated from SR QU-D5)                             and accident sequences, as discussed in QU-D5 for the FPRA.
Revision 0                                                                                                                                                           Page V-16

0 Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR           Basis and Recommendation                    Disposition in Fire PRA Update 1-26           By using the altered events table, and          PRM-A2         Given the large number of events set to zero This F&O has been closed.
setting recovered events to 0.1 (scoping)       SY-Al 1        in the altered events report (2764), and the The action taken to address this item was and the other events in the fault tree to zero, SY-A15         large number of basic events greater than    specifically included in the focused-scope random failures that may fail the HEP are                      1 E-03 in the PRA, there are likely numerous Peer Review.
not included in the results. Ifthe random                      events not included in the model.
events are greater than 2 orders of                            Review the values set to 0.0 in the altered magnitude below the top or contribute to 1%                    events report and modify the solution to of the system failure rate, they should be                     include the random failures in the results, included based on SY-A15. (This F&O                            when applicable.
originated from SR SY-A15 and PRM-A2) 1-27           Significant fire compartment contributors to    FQ-E1          Requirement of LE-F1, F3.                    This F&O has been resolved.
LERF are report.
summary     documented    in Appendix However,           C of the the contribution  LE-F1          Document the contributors to LERF based      Added LERF top cutsets and importances from plant damage states is not provided or     LE-F2          on the requirements of LE-F1 of the internal run as well as sensitivity analysis in fre omlntributors d        stes is nouroded or oma                        Levents                section of the standard, as required Summary Report. Also performed and the contributors from LEB SRs. Sources of       L-F3           by FE-Q1. Document the Sources of            documented the uncertainty evaluation for uncertainty, including sensitivity analysis     UNC-Al         uncertainty, including sensitivity analysis  LERF.
performed, are not evaluated for LERF.                         performed for CDF in Appendix D of the (This F&O originated from SR LE-F1)                            Summary Report.
1-28           It appears in the analysis that MCC fires for   FSS-A4         See FAQ 042. Scenario 058C-A showed a        This F&O has been closed.
un-vented MCC fires are not considered to                      CDF of 1 E-05 prior to screening.            The action taken to address this item was damage targets outside the selected MCC.                       Include in the model large MCC fires         specifically included in the focused-scope See Scenario 058C-A. The original draft of                     propagating outside of sealed MCCs           Peer Review.
FAQ 42 included discussion on MCC fires, and provide a probability of the fire propagating outside the MCC. This recommended approach was not included in the final FAQ, and as a result, the treatment of MCC fires does not meet the guidance in the FAQ for a sealed-cabinet. The end result is that the FPRA should consider MCCs as unsealed, due to the possibility of energetic fires resulting in the MCC door being opened, or as a result of maintenance on the MCC being the cause of the fire (actual events) when the door is open at the start of the fire. It is our understanding that the NRC Fire PRA folks also do not consider MCCs to Revision 0                                                                                                                                                       Page V-17

Florida Power & Light                                                                                                                   Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR            Basis and Recommendation.                   Disposition in Fire PRA Update be sealed cabinets. The proposed industry approach basically summarizes to the following: MCCs open at the top are considered unsealed, but MCCs that are sealed at the top have around a 0.1 probability of propagation. The data analysis for this value was not validated, but was considered conservative at the time of the proposal, since several fires where it was not clear if they came outside of the MCC were assumed unknown (1/2 an event).
(This F&O originated from SR FSS-A4) 1-29           The control room abandonment scenario            FSS-B2          The 0.1 scoping analysis is not based on       This F&O has been closed.
106-A, uses a CCDP of 0.1 (scoping value),                       analysis of plant specific procedures or       The action taken to address this item was However, the scenario is one of the top CDF                      analysis. The HEP may be higher or lower,      specifically included in the focused-scope scenarios. (This F&O originated from SR                          depending on procedures. Additionally, the     Peer Review.
FSS-B2)                                                          circuit analysis may show the Remote Shutdown Panel may not function for some scenarios due to MSOs.
Provided detailed analysis for MCR abandonment CCDP.
1-3            The internal events PRA model has                AS-B1           As a result of assuming a reactor trip and     This F&O has been resolved.
numerous locations in the model where the        ES-Al           not mapping components/equipment to            The issues and concerns identified in the specific initiating event results in a model                     modeled internal initiating events; the risk   F&O related to the fire-induced initiating impact. For example, under gate U3QT07;          ES-A3           can be under-estimated. In this case, since    events were reviewed. The review found initiating events that can cause a PORV or       ES-A4           the general approach used is systematic,       several instances where a change to the SRV to lift are ANDed with the failure to        FQ-A2           this problem is difficult to determine without modeling was required to allow the existing reclose the PORV or SRV. In this case,                           significant effort to combine the impact of    treatment methodology to be retained. The special initiator %ZZIP6U3 is identified as an                   each modeled impact. In most cases, the        review did not identify any instances where initiating event that will cause a PORV lift,                    modeling results in non-conservatism in the    specific fire initiating event logic beyond that along with %ZZT2U3. Equipment that can                           result. However, the fix for feed-and-bleed    already in the model was needed.
cause each are not mapped or modeled in                          resulted in conservatism for most of the the Fire PRA. As a result of a previous                          scenarios where FW is not initially lost. In review, the modeling of Feed-and-Bleed was                       either case; whether modeled conservatively changed to assume a loss of feedwater (low                       or nonconservatively, the standard SG level) occurred. The shorter time results                     requirements in this area are to model the in a higher HEP for feed-and-bleed in all                        impact of the FPRA accurately.
scenarios, regardless of whether a loss of Revision 0                                                                                                                                                             Page V-18 0

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                      Fire PRA SR          Basis and Recommendation                      Disposition in Fire PRA Update FW occurred. However, numerous other                            Map all identified internal events initiating modeling impacts can occur, that are not                        events to the specific components that can modeled. Under gate 162115, logic for                           cause the event, and modify the FPRA to HVAC unit 3S230 failure to start is included                    determine the CCDP based on the fire-when a Loss of offsite power would occur.                        induced initiating event that results.
This logic is applicable only for when a LOOP occurs, and not applicable for non-LOOP events. This type of logic is contained throughout the internal events PRA modeling. Another example is under gate El 104A, where loss of DC power results in lockout relay failures. There are many other examples throughout the PRA. Additionally, the identification of the specific initiating event for quantification was not performed per the requirements of FQ-A2. For quantification, the modeled initiating event is assumed to be a reactor trip in all cases.
This treatment does not meet the intent of SR FQ-A2, where the quantified model should encompass the risk contribution from all applicable initiating events.
1-30          Fire modeling was conducted via generic fire     FSS-C1          The present analysis provides a bounding      This F&O has been closed.
modeling from which Zones-Of-Influence           FSS-G1          approach for fire severity in most cases. The action taken to address this item was (ZOI) for specific initiator types was                           Perform 2-point fire modeling, when           specifically included in the focused-scope generated. The ZOls were used to define                          applicable, for significant fire scenarios. Peer Review.
bounding fire characteristics for each fire scenario. Characteristics that are used to bound potentially risk contributing fire events are identified in Attachment B of the Fire Scenario Report, (Report 0493060006.004).
Based on the use of a bounding approach this SR is judged to be met at CC i.
Significant fire scenarios should be developed with 2-point fire modeling. (This F&O originated from SR FSS-C1)
Reviion                                                                                                               Pag V-I Revision 0                                                                                                                                                        Page V-19

Florida Power & Light                                                                                                                     Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR             Basis and Recommendation                      Disposition in Fire PRA Update 1-31           Fire scenario evaluation tools were               FSS-C2         Significant scenarios are not developed          This F&O has been closed.
developed based on the Generic Fire               FSS-G1         considering fire growth. The current             The action taken to address this item was Modeling Treatments (Project Number                              approach is conservative.                        specifically included in the focused-scope SPH02902.030). These                                             Include fire growth times for significant fires, Peer Review.
walkdown/evaluation tools are based on                           where growth time is available in bounding fires that are assumed to cause                         NUREG/CR-6850 or FAQ-052 for transient target damage at a height above the base                         fires..
fire with the fire burning at peak intensity.
Because these tools assume a fire burning at peak intensity this SR is considered met at CC I. (This F&O originated from SR FSS-C2) 1-32           No evidence was identified that suggests          FSS-C3         Significant scenarios do not consider decay. This F&O has been closed.
that fires were assumed to burnout over a         FSS-G1         Include growth and decay for significant fire    The action taken to address this item was period of time. Evaluation/scoping was                           scenarios in the FPRA.                           specifically included in the focused-scope estimated using a peak heat release rate as                                                                       Peer Review.
dictated in Attachment B of the Fire Scenario Report, (Report 0493060006.004).
Accordingly this SR is considered not met for CCII/Ill. (This F&O originated from SR FSS-C3) 1-33           Except for the MCR fire scenarios, no other       FSS-D8         Systematic.                                      This F&O has been closed.
fire scenario has used the Non-Suppression                       Apply non-suppression factors for significant    The action taken to address this item was Probability (NSP) in PTN fire model at this                      fire scenarios, when applicable.                 specifically included in the focused-scope time. (This F&O originated from SR FSS-                                                                           Peer Review.
D8) 1-34           No evidence was found that supported              FSS-G4         Systematic issue.                                This F&O has been resolved.
confirmation of conformance of fire rated                        Provide the documentation that supports          The treatment of barriers in the MCA is barrier segments to applicable test                              confirmation of conformance of fire rated        based on information in the Fire Hazards standards. Additionally, the effectiveness,                      barrier segments to applicable test              Analysis and supplemented with walkdown reliability, and availability of any passive fire                standards, and the barrier effectiveness,        observations. The analysis documentation barrier feature credited does not appear to                      reliability and availability,                    was updated to provide this information.
be performed. (This F&O originated from                                                                           The MCA was modified as needed to SR FSS-G4)                                                                                                        incorporate the results of this effort.
Revision 0                                                                                                                                                              Page V-20 0

Florida Power & Light                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update 1-35           The multi-compartment analysis assumes a        FSS-G5          Systematic issue                                This F&O has been closed.
bounding value of 7.4E-3 for evaluation of                      For any scenario selected if the adjoining      The action taken to address this item was active fire barrier elements. Actual fire                       physical analysis units are separated by        specifically included in the focused-scope barrier elements are not considered instead                     active fire barrier elements, UANTIFY the       Peer Review.
the failure probability of a fire door is                       effectiveness, reliability, and availability of assumed for active barrier element failure                      the active fire barrier element.
because this failure probability represents the highest single probability of a barrier failure. Accordingly this analysis provides a qualitative bounding assessment fire barrier feature failure probability. (This F&O originated from SR FSS-G5) 1-36           HFEs included in the altered events report      FQ-C1           Systematic Issue                                This F&O has been closed.
are not treated under dependency                HRA-C1          Perform dependency analysis for HFEs in         The action taken to address this item was evaluations Even though the values are          HRA-D2          the Altered Events Table, and provide a         specifically included in the focused-scope screening values, the dependency                                quantification process that incorporates the    Peer Review.
evaluation may result in a higher HEP,          HR-H3           new HFE dependency.
especially if more than 2 events are in a       QU-C1 single cutset. The sensitivity study case 3     QU-C2 documented in Appendix D of FPRA Summary Report 0493060006.005 shows that doubling all non-recovery HEPs using multipliers greater than 1 yields a delta CDF increase of 6.60E-5, or 24.7% of the base fire CDF. Furthermore, if doubling the HEPs increased CDF by -25%, it stands to reason that halving the same set of HEPs would decrease the CDF by a similar amount. It is reasonable to assume that a detailed analysis could reduce most of these screening HEPs by at least half, and in many cases by much more. (This F&O originated from SR FQ-C1)
Page V-21 Revision 00                                                                                                                                                         Page V-21

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR            Basis and Recommendation                   Disposition in Fire PRA Update 1-37           Significant contributors to Fire PRA results    FQ-E1           Requirement of QU-D7                         This F&O has been resolved.
are included in Section 4.3 and the             QU-D7           Provide importance measures as required      Importance measures for CDF and LERF appendices of the Summary Report. This                          by QU-D7 and FQ-E1                           have been determined and added to the includes a list of operator actions that                                                                     Summary Report.
contribute to CDF. However, no importance measures are provided for CDF or LERF.
(This F&O originated from SR QU-D7) 1-38           Results of the Fire PRA did not include the     FQ-F1           Systematic Issue                             This F&O has been resolved.
following: (e) the total plant CDF and          QU-F2           Provide required documentation per QU-F2     The documentation of the analysis results contributions from the different initiating     UNC-A2          and FQ-Fl.                                   has been expanded to include the events and accident classes (i) the                                                                          information noted in the F&O. These results uncertainty distribution for the total CDF ()                                                                were also reviewed for reasonableness and importance measure results (1)asymmetries                                                                    no issues or concerns were identified.
in quantitative modeling to provide application users the necessary understanding of the reasons such asymmetries are present in the model (m) the process used to illustrate the computer code(s) used to perform the quantification will yield correct results process. Some of these issues are listed in other F&Os.
However, item e(accident classes), I (asymmetries) and m (validation of computer codes) is not covered elsewhere. (This F&O originated from SR QU-F2) 1-4            Fire-Induced Spurious PORV opening is           ES-B3           Systematic issue of how the PORV, other      This F&O has been closed.
modeled under Gate 3S2ORFIREEQUIV               PRM-A4          small-small LOCA initiating events and small The action taken to address this item was (2/6 gate). However, this small LOCA impact     PRM-B9          LOCA lEs are modeled in the PRA.             specifically included in the focused-scope is only added to three locations in the PRA                     For any fire-induced initiating events,      Peer Review.
model, while small LOCA is located in 11                        including PORV opening and other small locations in the CAFTA Model. For example,                      LOCA initiating events, ensure the newly the fire-induced small-loca logic is not                        developed logic is added to all locations in included under gate GHL01 or 7 other                            the PRA model to ensure the PRA model locations. For small-small LOCA, the PORV.                      solves correctly. Move this F&O to PRM, Similarly, small-small LOCA is included in 9                    when evaluated. Also affects the equipment locations in the PRA model, while the Fire-                     selection, only with regard to the impact of Induced PORV logic (and other small-small                       the equipment on the modeled initiating logic) is only included in 3 locations. The Revision 0                                                                                                                                                       Page V-22

Florida Power & LIqht                                                                                                                    Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update above are examples for two of the reviewed                       event.
added logic. The fire-induced initiator event fault trees have been added into the PRA model. The consistency of the model changes have been reviewed. One question about the changes is that the fire risk could be potentially underestimated. For example, gate 3FIRES2 has been added to simulate the small LOCAs induced by fires. Under gate 3S2ORFIREEQUIV, gate 3FIRES2 and internal event small LOCA initiator
               %ZZS2U3 are ORed, which seems to be appropriate. However, %ZZS2U3 is under 11 parent gates, while 3FIRES2 only has one parent gate. If the intended fire damage is small LOCA, all system functions affected by small LOCAs should be affected. (This F&O originated from SR ES-B3) 1-40          The quantification of significant basic          FQ-F1           Requirement of QU-F6 and FQ-Fl.                 This F&O has been resolved.
events, cutsets and accident sequences is        QU-F6           Provide the quantification of significant basic The Summary Report has been updated to not provided. Additionally, the definitions      UNC-A2          events, cutsets and accident sequences,         provide the importance measures of the used for significant basic event, significant                    and the definition used for significant basic   model basic events, top 90% of all plant cutset, and significant accident sequence                        event, significant cutset, and significant      cutsets, and a review of the scenarios are not provided. (This F&O originated from                      accident sequence                               contributing more than 1% of the total risk.
SR QU-F6) 1-41          The HRA does not look at the Fire Specific       HRA-D2          The simplified factors included in the HEP      This F&O has been closed.
factors affecting: (a) quality [type (classroom  HR-H2           modifier approach does not provide the          The action taken to address this item was or simulator) and frequency] of the operator     PRMB1 1         equivalent of a detailed HRA.                   specifically included in the focused-scope training or experience (b) quality of the                        Provide a detailed HRA for significant HEPs     Peer Review.
written procedures and administrative                            in the FPRA results. Detailed HRA should controls (c) availability of instrumentation                     account for the Fire-Specific factors as listed needed to take corrective actions (d) degree                     above. The present multiplier method does of clarity of cues/indications (e) human-                        not appear to meet the requirements of the machine interface (f) time available and time                    standard for detailed HRA.
required to complete the response (g) complexity of the required response (h) environment (e.g., lighting, heat, radiation) under which the operator is working (i)
Revision 0                                                                                                                                                          Page V-23

Florida Power & Light                                                                                              Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR          Basis and Recommendation  Disposition in Fire PRA Update accessibility of the equipment requiring manipulation () necessity, adequacy, and availability of special tools, parts, clothing, etc. As a result, the HRA does not appear to meet the PRA standard requirements for a detailed HRA. (This F&O originated from SR HR-H2)
Page V.24 Revision 00                                                                                                                                 Page V-24 0

0 Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                      Fire PRA SR           Basis and Recommendation                     Disposition in Fire PRA Update 1-44           The FPRA models 0.1 HEP values for              HRA-Cl          Appears to be a significant non-conservative  This F&O has been closed.
several recovery actions that are not in the    PRM-B131        impact to PRA results. Given the actions in   The action taken to address this item was SSA and are not in the fire safe shutdown                       the altered events report are being added to  specifically included in the focused-scope procedures (See F&O 6-11). These recovery                       the model as needed recoveries in order to    Peer Review.
actions are input into the PRA model by                         ensure risk is low, and given the resulting adjusting the random independent failure                        recovery actions do not show up in the probability for the recovered component to                      results in most cases, there appears to be a 0.1 and the logical true setting for                            disconnect between the addition of new components that prompted the recovery to                        actions to the procedures and the 0.0. This approach introduces several                           quantification of these actions in the FPRA.
issues, including the following: The use of                     It appears part of the disconnect is that the 0.0 values that are intended to account for                      logic modeling, as modified by the altered recovered equipment eliminates the 0.1                          events table, results in the recovery values recovery HEPs altogether in several                             being screened from the results.
instances. For compartment 070-AB, basic                        Revise the approach described for assigning event EC8R330303 is set to 0.1 for recovery                      recovery HEPs via the altered events table of alternate feed to load center 3H, but this                    and ensure the quantification produces the recovery feeds into AND gate E3013H. This                       intended results.
AND gate appears to be nullified, however as the second input to the gate is false due to the 0.0 values present in altered events (based on visualization of settings from the altered events file for 070-AB in CAFTA). A quantification of 070- AB produced 10,000 cutsets and no instance of EC8R330303 was present. A further review of the 070-AB quantification indicated the following events placed in the altered events table as recoveries do not appear in the final cutsets:
ATPXPUMPASTRT, EB2F33003H, ECBR33AA15, FAVC3-1606, MAVX3-311, MSVR3-31 1. Further review of other areas and other recovery values (0.1 and 0.0 values in the altered events report) appears to indicate similar problems will occur. The above are examples (potential issues),
which appear to be logic problems resulting from the use of 0.0 and 0.1 inputs in the altered events report. Additional problems are likely for other events. (This F&O originated from SR HRA-C1)
Page V.25 Revision 00 Revision                                                                                                                                                          Page V-25

Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR           Basis and Recommendation                    Disposition in Fire PRA Update 1-45           The method and calculations for transient       FSS-D3         The Severity factors used are in significant This F&O has been closed.
fire severity factors (SF) are not clearly                     fire scenarios in the FPRA.                  The action taken to address this item was documented and several SF values in                            Document the severity factors used for each  specifically included in the focused-scope FRANC model are not consistent with the                        scenario including the basis. Revise the     Peer Review.
ones listed in the FSS report. For example,                    transient severity factors to remove double fire scenarios 079A-J/K/L in Appendix A of                     counting of the area factor included in both FSS report have an SF value of 3.05E-2,                        the square footage of the compartment, and which is used in the FRANC model, but is                       the 8E-02 calculated in Section 8.4 of the not consistent with the 8E-2 value included                    FSS report.
in FSS report section 8.4, GENERAL TRANSIENT SEVERITY FACTORS.
FPL/ERIN staff reviewed this issue and stated that the SF is calculated based on a floor area factor (FSS report section 8.3) times the 8E-2 transient ignition frequency adjustment factor (FSS report section 8.4).
For 079A-J/KIL, this method results in a factor of 1.52E-2. A factor of 3.05E-2 was used in FRANC. The above points to the following: a) The severity factors used are not well documented, traceable, or consistent with what is provided in the text of the FSS report. B) The two severity factors basically double count the area severity factor. For the last case (B), the 8E-02 already includes a consideration, based on experience, that the fire that occurred is near a target (component). As a result, the 6 events listed just happened to not be near any components, resulting in a low probability of damage. To put this another way, it can not be demonstrated that the 8E-02 factor is due to the small size of the transient fires rather than being due to the fire being in a location not near a component. A review of the control/aux bId transient fires was performed. The first event caused an automatic suppression system actuation, which indicates a fairly large fire occurred. It takes a pretty good size fire to Page V-26 Revision Revision 0                                                                                                                                                      Page V-26

Florida Power & Liqht                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                      Fire PRA SR           Basis and Recommendation                      Disposition in Fire PRA Update raise the fire detectors up to 160+ degrees.
A second event is described as "A leaking regulator ignited leaking propane." This can obviously be a larger fire. Since the location/area of the originating fire is in the 8E-02 factor, and in the "area" factor, the double counting results in an underestimation of the likelihood of fire damage for a transient fire. (This F&O originated from SR FSS-D3) 1-46           The PTN FPRA methodology generally does         FSS-D9          Requirement of FSS-D9                          This F&O has been closed.
not include postulation or evaluation of                        Perform a qualitative evaluation of smoke      The action taken to address this item was smoke damage. Additional review shows                           damage to FPRA equipment per the               specifically included in the focused-scope that the smoke issues do not affect the                          requirements of FSS-D9.                        Peer Review.
FPRA results significantly. However, the FPRA does not include a qualitative evaluation of smoke damage to FPRA equipment. (This F&O originated from SR FSS-D9) 1-5            HEP EHFCLR3BH is included in the FPRA           HRA-C1          Completeness issue in the documentation of     This F&O has been closed.
model, without specific analysis in the task                    the HRA.                                       The action taken to address this item was 12 report. This is a modified HEP from the                      Ensure documentation of all HEPs is            specifically included in the focused-scope internal events PRA. Additionally,                              provided in the HFE report.                    Peer Review.
MHFP3BAMT is included in the model as a screening value without documentation in Task 12. (This F&O originated from SR HRA-C1) 1-6           Although many of the instruments affecting       ES-Cl           Systematic issue. As a result of this, the     This F&O has been closed.
operator actions are included in Table A-1 of    HRA-A3          equipment list for the FPRA (considered        The action taken to address this item was the HRA report, and many are traced, none                        separate for this review) does not include     specifically included in the focused-scope of the instruments reviewed (partial review)                     any of the identified instruments in Table A-1 Peer Review.
appear to have been tagged as FPRA                               of the HRA.
affecting. As a result, the instruments do not                   Include needed Instruments in the PRA appear to be in the FPRA equipment                               equipment list, for later modeling of the selection process or equipment list. For                         impact on the FPRA HEPs.
example, under GHFPS1 RCRC, FI-3-943 is listed and included in the SSA. However, Revision 0                                                                                                                                                         Page V-27

Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                   Fire PRA SR            Basis and Recommendation                      Disposition in Fire PRA Update this is listed as non-PRA affecting as is not in the TAGBE, UNL table or other associated table.
(This F&O originated from SR ES-Cl).
Additionally, when new equipment is added that are not in the SSE (for example; LOCAL CHARGING PUMP HYDRAULICA COUPLING TEMPERATURE TI-3/4-6716/6717/6718), these do not appear to be traced or included in the PRA equipment list.
Most of the PRA (non-SSE) instruments are not identified by number.
1-9            Recovery to the Charging Pump Suction          CS-Al1          1st item does not appear to be significant. This F&O has been closed.
valve 114A and the associated components       CS-C3           SSA verifies MOV 350 is available, but the     The action taken to address this item was is provided through the excluded events                        documentation is very difficult to follow. specifically included in the focused-scope table and the altered events table.            LE-A2           However, the MOV and the associate             Peer Review.
However, the recovery is actually crediting    PRM-B15         components do not show up in the flow through MOV-3-350. This flow path is                      Equipment List for the PRA, nor will it show not included in the FPRA model. Similarly,                     up in the importance lists, results, etc. The the Normal Containment Coolers are not                         surrogate event (0.1 applied to 114A) does modeled in the FPRA, but recovered in the                      not represent what is in the model.
altered events table through G174 0.1                          Additionally, since the flow path was not recovery. The above are examples. The                          modeled, the FPRA considerations may not entirety of the altered events table has not                   all be considered. 2nd item: This one may been reviewed for recoveries that may be                       be significant, since the containment cooler bringing in additional components not in the                   can be complicated, and a likely mismatch FPRA or FPRA Equipment List (This F&O                          between the SSA and the modeled FPRA originated from SR CS-Al1)                                     logic could result in significant differences.
Add the MOV-3-350 and the associated flow path to the PRA model, normal containment coolers. Additionally, review the altered events report to determine if additional credited flow paths are modeled, where the actual components and support logic is not in the FPRA.
2-1            The plant partitioning task does not include   PP-B1           Section 3.11.5 of FHA states that man-hole     This F&O has been resolved.
detailed discussion with respect to this SR    PP-B7           covers are justified as three-hour fire        Walkdowns of fire zone boundaries were PP-B7. The manholes are modeled as                             boundary although they need not to be Revision 0                                                                                                                                                        Page V-28

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR            Basis and Recommendation                     Disposition in Fire PRA Update separate fire compartments. However, no                        specifically rated as fire barrier. Therefore, performed and documented in support of a walkdown for these manholes has been                            the modeling of manhole as fire                review of the Fire Hazards Analysis update.
performed. No justification for the modeling                   compartments is considered acceptable          Additional discussion regarding the basis for approach has been provided except being                        although no walkdown has been performed        the ignition frequency for the manholes was briefly mentioned in Section 2.2 of Report                     for the manholes. Other credited barriers are  added to the documentation.
PTN-PSA-7.01 Revision 2. Walkdowns were                        discussed in PP-B2-4 above also not documented for spatial separation                     Consider adding justification for the or other boundaries that are not fire rated                    modeling of manholes according to the but was credited in the FPRA. (This F&O                        requirements in SR PP-B7. Consider originated from SR PP-B7)                                      performing walkdown for manholes with significant risk contribution. Also, document walkdowns on all credited, nonrated barriers credited in the FPRA.
2-10          A review of the HRA report and the recovery      PRM-B11        One of the identified issues generates non-    This F&O has been closed.
rule file used in the FRANC model shows                        conservative results and the other one is      The action taken to address this item was that the majority of the HEPs and HFE                           conservative. The second example is likely     specifically included in the focused-scope combinations were treated properly in the                       significant.                                   Peer Review.
PTN fire PRA model. However, isolated                          Consider reviewing the recovery rule file for cases indicate the following issues: 1. Some                    consistency against the HRA report HEPs were not applied (or documented)                          documentation. Also consider updating the properly. For example, EHFPDOSTXT,                             HRA combination evaluation.
Failure to cross-connect unit diesel oil storage tanks to extend availability of fuel for EDG, was supposed to be set to 1.0 per Table A-2 in HRA report page A-19.
However, it is set to 2.3E-2 in the recovery rule (in both the HRA report and actual rule file). A review by FPL/ERIN staff showed that the example HEP was included in the UNL table (failed for all scenarios), however, it was also included in the Excluded Events table for multiple fire scenarios. Since this HFE has no cue in the main control room, it was intended to fail this HEP for all fire scenarios. Therefore, the Excluded Events table should be updated. 2. A review of the top cutsets in fire sequence 096A show the HFE combination CHFPSTPRCP and GHFPINJVLVS has not been considered in Revision 0                                                                                                                                                        Page V-29

Florida Power & Light                                                                                                                 Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR            Basis and Recommendation                     Disposition in Fire PRA Update the HRA evaluation. FPLIERIN staff concurred that some HEP combinations may be missed, which render conservative results. Because fire scenarios for fire compartment 096 will be refined for realism, the HEP combinations are expected to be re-visited. (This F&O originated from SR PRM-B 11) 2-12           A comparison of basic events between the        PRM-B12        Isolated error. However, the error may be      This F&O has been closed.
internal event (Rev. 7) and fire models has     PRM-B13        significant                                    he action taken to address this item was been performed. For the new basic events                       Consider providing basis for the modeled       specifically included in the focused-scope (about 650 basic events are identified) in fire                new basic events in the fire PRA model.        Peer Review.
model, the majority is set to either 0 or 1,                   Correct AHFPAFWFLO probability in which simulates the fire impact in the fire                    FRANC.
scenarios. The ones with other values are checked and found to be lacking sufficient basis as required by SR PRM-B13. For the modified HEPs, the majority seems to be ok by updating with a more bounding value of 1.0. However, the following event is an outlier, which may result in early truncation of the cutsets with this AHFPAFWFLO, OPERATOR FAILS TO THROTTLE UP AFW FLOW, 1.7E-4. For the deleted basic events in fire model, all the changes have been traced in the PRM report. (This F&O originated from SR PRM-B13) 2-14           Fire compartment 096 is the top contributor     PRM-B9         Current analysis for top fire scenario 096A is This F&O has been closed.
for U3 CDF. Tracing the failed basic events /                  conservative. This issue applies to other fire The action taken to address this item was components /cables shows that the                              scenarios in the FPRA.                         specifically included in the focused-scope sequencer failures seem to contribute to loss                  Consider updating the top fire scenarios to    Peer Review.
of redundancies. Since the cables travel to                    remove conservatism related to sequencer rooms housing redundant trains, the                            modeling and failures.
assumed failure of cables fail the sequencer, which in turn fails the switchgears. For example, circuit analysis for 3X03-NPO-3AA states, "ADDED CABLES IN SEQUENCER THAT CAN PREVENT SUT FROM Revision 0                                                                                                                                                        Page V-30 0

0                                                                                                                                                            0 Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION-OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR            Basis and Recommendation                   Disposition in Fire PRA Update POWERING SWGR THRU 3AA05 BREAKER." A discussion with FPLERIN staff indicated that the analysis for fire compartment 96 is not yet complete as documented in the summary report. It is anticipated that the approaches and refinements used for other plant locations will result in more realistic risk results for the room. On the other hand, FPL staff also identified that there is an open item associated with fire zone 096 (SSA-3GG-13). (This F&O originated from SR PRM-B9) 2-15           Most new events are added to the model in       PRM-B13         Systematic Issue.                            This F&O has been closed.
order to assess spurious operation, and                         Consider adding to the documentation         The action taken to address this item was other Fire PRA effects. However, there is no                    whether any events added to the fire PRA in  specifically included in the focused-scope documentation supporting the events, and                        Table D-1, 2, or 3 of the component          Peer Review.
as such, they do not meet the DA                                selection report are new to the PRA. If new, requirements as referenced in PRM-B13.                          add the details of the event to the This SR lists an exception (DEVELOP a                           documentation per the associated standard defined basis to support the claim of non-                      requirements or provide justification to the applicability of any of these requirements in                   PRA documentation of non-applicability of Section 2), which is not provided. (This F&O                    any DA SR.
originated from SR PRM-B13) 2-16           RCP Spurious start logic under                  PRM-B9          Isolated issue. Risk impact is not known     This F&O has been closed.
3FIRERCPSPUR3 appears to be incorrect.                          Consider refining the model for the spurious The action taken to address this item was First, the spurious start of an RCP typically                   start of RCPs. Also address the fire impact  specifically included in the focused-scope takes 2 spurious operations, including start                     of the operator action to trip the RCPs. Peer Review.
of the lift pump, and then start of the main pump. Second, the logic does not include any consideration of whether the operator already tripped the pumps, and they restarted or whether the RCP received a spurious signal, and then operator trip of the pump is not possible (may be possible with a single spurious depending on the design).
Finally, if the RCP restarts (as modeled),
tripping of the RCPs may not be possible Revision 0                                                                                                                                                       Page V-31

Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR             Basis and Recommendation                  Disposition in Fire PRA Update depending on the operator actions performed. This consideration needs to be included in the determination of the event for operator trips the RCPs. Under gate 3FIRERCPSPUR1, 3FIRERCPSPUR1 (RCP Seal Failure Following Spurious Pump Start Due to Fire) is AND'ed with the HEP CHFPSTPRCP, which is not failed by any fire scenario. This event is evaluated in the HRA report, but the fire impact for an operator to trip the RCPs is not evaluated, (This F&O originated from SR PRM-B9) 2-18           PTN LERF model development is                  LE-C 1          Systematic issue.                            This F&O has been closed.
documented in the Component and Cable          LE-C3           Identify and document any new accident       The action taken to address this item was Selection Report 0493060006.001, Rev. 1,                       progression beyond the onset of core         specifically included in the focused-scope Section 4.5, LARGE EARLY RELEASE               LE-C8 damage that would be applicable to the Fire  Peer Review.
FREQUENCY (LERF). This section                 PRM-B1          PRA that were not addressed for LERF described the excluded LERF sequences in       PRM-B14         estimation in the Internal Events PRA.
internal events model. No discussion is                        REVIEW significant accident progression identified on any new accident progression                     sequences resulting in a large early release beyond the onset of core damage that would                     to determine if repair of equipment can be be applicable to the Fire PRA that were not                    credited. JUSTIFY credit given for repair addressed for LERF estimation in the                           (i.e., ensure that plant conditions do not Internal Events PRA. As a result, significant                  preclude repair and actuarial data exists accident progression sequences resulting in                    from which to estimate the repair failure a large early release have not been                            probability [see SY-A24, DA-C15, and DA-reviewed to determine if repair of equipment                   D8]). INCLUDE accident sequence can be credited for the FPRA LERF models                       dependencies in the accident progression
[LE-C3 CC-lI requirement]. In addition,                        sequences in a manner consistent with the accident sequence dependencies in the                          applicable requirements of 2-2.2, as accident progression sequences have not                        appropriate for the level of detail of the included in a manner consistent with the                       analysis [LE-C8 requirement].
applicable requirements of 2-2.2, as appropriate for the level of detail of the analysis [LE-C8 requirement]. (This F&O originated from SR PRM-B14)
Page V-32 Revision 00                                                                                                                                                     Page V-32

Florida Power & Light                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                   Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update 2-19          The PTN fire LERF model is built upon a         LE-Al          Draft internal event LERF model is used.         This F&O has been closed.
draft internal events LERF model (rev. 8),      LE-A2          Changes to the draft model when finalized        The action taken to address this item was which needs to be updated when that model       PRMB14         could be significant.                            specifically included in the focused-scope is finalized. (This F&O originated from SR     PRM-1          Update the fire LERF model when the              Peer Review.
PRM-B14)                                       PRM-B15        internal event LERF model is finalized.
2-22           Beyond the Generic Fire Modeling               FSS-C6         Without detailed fire modeling for significant   This F&O has been closed.
Treatments, the Fire PRA did not include        FSS-D3         fire scenarios, the results are conservative. The action taken to address this item was additional detailed fire modeling for most fire FSS-G1         Consider performing additional detailed fire     specifically included in the focused-scope compartments. (This F&O originated from                        modeling for target damage timing when the       Peer Review.
SR FSS-C6)                                                     exposure environment exceeds the damage threshold.
2-26          The system unavailability records for the       FSS-D7         Systematic issue. The intent for Capability      This F&O has been closed.
plant have not been reviewed in crediting       FSS-D8         Category II is to additionally require a review  The action taken to address this item was fire detection and suppression systems.                        of plant records to determine if the generic     specifically included in the focused-scope (This F&O originated from SR FSS-D7)                           unavailabilitycredit is consistent with actual   Peer Review.
system unavailability. Outlier experience would be any experience indicating that actual system is unavailable more frequently than would be indicated by the generic values.
Consider performing and document the review of plant records to determine if the generic unavailability credit is consistent with actual system unavailability. Outlier experience would be any experience indicating that actual system is unavailable more frequently than would be indicated by the generic values.
2-29          PTN FSS report 0493060006.004 Revision          FSS-D10        Documentation issue. However, the                This F&O has been closed.
1, App. A documents the SCENARIO                FSS-D1 1       inadequacy of the walkdown documentation         The action taken to address this item was

==SUMMARY==
REPORT, which includes the                             cannot provide detailed information for          specifically included in the focused-scope combinations of fire sources and target sets. FSS-H10        scenario development or detailed fire            Peer Review.
However, the walkdown documentation is                         modeling.
lacking. Currently, the walkdown results are                   Consider enhancing the process and documented directly in the FRANC                               documentation of the source-target data Revision 0                                                                                                                                                         Page V-33

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                   Fire PRA SR             Basis and Recommendation                     Disposition in Fire PRA Update database. Consider a walkdown                                 collection walkdowns. See recommended documentation package which would include                     walkdown information in NUREG/CR-6850.
elements such as a data collection procedure, documentation of who performed what walkdowns on what dates, documentation that review of the collected source-target data was performed, etc. (This F&O originated from SR FSS-D10) 2-31           The PTN fire PRA model has not completed      FSS-H8          Quantitative results for the identified        This F&O has been closed.
the quantitative results for any scenarios                    significant fire scenarios in the multi-       The action taken to address this item was analyzed quantitatively in a manner that                      compartment analysis should be generated       specifically included in the focused-scope facilitates Fire PRA applications, upgrades,                  and documented.                                Peer Review.
and peer review. (This F&O originated from                    Quantify the identified significant fire SR FSS-H8)                                                    scenarios in the multi-compartment analysis and complete documentation.
2-37           The fire PRA has not identified how the       LE-A3           Requirement not met.                           This F&O has been closed.
physical characteristics (unique to fire      PRM-B1 5        Identify how the physical characteristics      The action taken to address this item was scenarios, or affected by fire scenarios)                     (unique to fire scenarios, or affected by fire specifically included in the focused-scope identified in LE-Al and the accident                          scenarios) identified in LE-Al and the         Peer Review.
sequence characteristics identified in LE-A2                  accident sequence characteristics identified are addressed in the LERF analysis. FPRA                      in LE-A2 are addressed in the LERF impact can affect accident sequences for                      analysis.
LERF such as failing containment isolation, affecting containment cooling or instrument air to containment, opening pressurizer PORVs (pressurizing containment), etc.
(This F&O originated from SR LE-A3) 2-38           The fire PRA did not re-visited the plant     LE-A5            Requirement not met.                          This F&O has been closed.
damage states defined in the internal events  PRM-B15         Re-define the plant damage states in the       The action taken to address this item was LERF model. (This F&O originated from SR                      internal events LERF model to account for      specifically included in the focused-scope LE-A5)                                                        any fire-specific characteristics              Peer Review.
2-39           The MSO Review List reviewed by the           LE-B1           Systematic issue.                              This F&O has been closed.
Turkey Point Expert Panel should have         PRM-B1 5         Use a systematic process and document the     The action taken to address this item was captured most of the requirement in this SR.                  identification of the fire-specific LERF       specifically included in the focused-scope However, a systematic process and                             contributors from the set identified in AMSE   Peer Review.
Revision 0                                                                                                                                                       Page V-34 0

0 Florida Power & Light                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                  Fire PRA SR             Basis and Recommendation                       Disposition in Fire PRA Update documentation are not available with respect                   standard Table 2-2.8-9.
to the identification of LERF contributors from the set identified in AMSE standard Table 2-2.8-9. (This F&O originated from SR LE-B1) 2-4            The FRANC model itself cannot generate        PRM-A3          This finding mainly focuses on                   This F&O has been closed.
total CDF / LERF and risk importance                           documentation. However, the unavailability      The action taken to address this item was measures. The models show the                                  of the cutset files and risk importance         specifically included in the focused-scope contributions from each quantified fire                        reports prevents the detailed analysis in       Peer Review.
sequences. As shown in the summary report                      other tasks such as HRA and circuit 0493060006 Rev 1, the Unit 3 CDF top                           analysis, etc.
cutsets are listed. However, the Unit 4 CDF,                   Document cutset files for U3 LERF and U4 Unit 4 CDF & LERF, the risk importance                        CDF/LERF and risk importance reports for values for each basic events are not                           each.
included. (This F&O originated from SR PRM-A3) 2-40          A separate FPRA LERF package is not            LE-C2            LE-C2 CC-Il requirement                         This F&O has been closed.
available. SR LE-C2 is assigned as not met     PRM-815          INCLUDE REALISTIC treatment of feasible         The action taken to address this item was since the LERF specific operator actions is                    operator actions for LERF-Specific HEPs (if      specifically included in the focused-scope not evident in the HRA report and the screen                  any) following the onset of core damage          Peer Review.
values are used for numerous operator                          CONSISTENT WITH APPLICABLE recovery actions for the fire-induced                          PROCEDURES.
component failures. (This F&O originated from SR LE-C2) 2-42          FPRA specific significant accident             LE-C4           SR LE-C4/5/6 Requirements                        This F&O has been closed.
progression sequences resulting in a large     LE-C5           For FPRA LERF model, evaluate the fire-          The action taken to address this item was early release have not been developed. The     LE-C6           specific aspects for the following               specifically included in the focused-scope FPRA LERF changes are directly                                 requirements in SRs LE-C4 through                Peer Review.
incorporated in the quantification fault tree. PRM-B15         CriNCLUDE model logic necessary to FPRA specific realistic generic or plant-                      provide a realistic estimation of the specific analyses for system success criteria                  significant accident progression sequences for the significant accident progression                       resulting in a large early release. INCLUDE sequences have not been developed. The                         mitigating actions by operating staff, effect of FPRA LERF changes are directly                                 fission product scrubbing on radionuclide incorporated in the quantification fault tree.                 release, and expected beneficial failures in FPRA specific system models that support                       significant accident progression sequences.
the accident progression analysis have not                     PROVIDE technical justification (by plant-Revision 0                                                                                                                                                         Page V-35

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR           Basis and Recommendation                    Disposition in Fire PRA Update been developed. The FPRA LERF changes                         specific or applicable generic calculations are directly incorporated in the quantification               demonstrating the feasibility of the actions, fault tree. (This F&O originated from SR LE-                  scrubbing mechanisms, or beneficial C6)                                                           failures) supporting the inclusion of any of these features. USE appropriate realistic generic or plant-specific analyses for system success criteria for the significant accident progression sequences. USE conservative or a combination of conservative and realistic system success criteria for non-risk significant accident progression sequences.
DEVELOP system models that support the accident progression analysis in a manner consistent with the applicable requirements for 2-2.4, as appropriate for the level of detail of the analysis.
2-43           The significant accident progression            LE-Cl0        LE-C10/12 CC-Il requirement                    This F&O has been closed.
sequences resulting in a large early release    LE-C12        REVIEW significant accident progression        The action taken to address this item was have not been reviewed to determine if          PRM 815       sequences resulting in a large early release   specifically included in the focused-scope engineering analyses can support continued           -        to determine if engineering analyses can       Peer Review.
equipment operation or operator actions                       support continued equipment operation or during accident progression that could                        operator actions during accident progression reduce LERF. The significant accident                         that could reduce LERF. USE conservative progression sequences resulting in a large                    or a combination of conservative and early release have not been reviewed to                       realistic treatment for non-significant determine if engineering analyses can                         accident progression sequences. REVIEW support continued equipment operation or                      significant accident progression sequences operator actions after containment failure                    resulting in a large early release to that could reduce LERF. (This F&O                             determine if engineering analyses can originated from SR LE-C10)                                    support continued equipment operation or operator actions after containment failure that could reduce LERF. USE conservative or a combination of conservative and realistic treatment for non-significant accident progression sequences 2-44           Uncertainty Evaluations (Sensitivity studies)   QU-E4         QU-E4 requirements.                            This F&O has been resolved.
should be performed for both CDF and            UNC-Al        Perform sensitivity studies should be          Parametric uncertainty and sensitivity has LERF model for Units 3 and 4 since the Revision 0                                                                                                                                                       Page V-36

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR             Basis and Recommendation                   Disposition in Fire PRA Update model uncertainties may have different          UNC-A2          performed for both CDF and LERF model for    been performed for CDF and LERF for both impact to specific model due to differences                     Units 3 and 4.                               Units. The results do not indicate any in plant designs, FPRA model details, and                                                                    change in the selection of parameters or etc. (This F&O originated from SR QU-E4)                                                                     assumptions are necessary.
2-6            Conservative screening values were used         FQ-C1           To meet CC-Il requirements, the risk         This F&O has been closed.
for risk-significant human actions, including   HRA-C1          significant HFEs should be evaluated in      The action taken to address this item was both the new fire specific safe shutdown                        more detail, as specified in HRA-C1. The     specifically included in the focused-scope actions identified and the non-fire actions     HRA-D2          approach used to estimate HEPs for the       Peer Review.
that exist in the internal events PRA and        HRA-E1          risk-significant, new fire-related operator included in the Fire PRA. It is noted that the  HR-G1           actions should use an approach that HEPs derived by the multiplier approach for     HR-G2           addresses both failure in cognition as well the non-fire actions are still considered as                     as failure to execute.
the screening values until a detailed analysis  HR-G3           Update the HRA by performed detailed HRA is performed. There are numerous                HR-l1           for all risk-significant HEPs, including the examples of significant HEPs in the results,    HR-12           HEPs analyzed using the multiplier method as well as significant screening HEPs set to                    in the HRA, as well as the screening HEPs 0.1 in the altered events report. The           PRM-B11         in the altered event report. The HRA estimation of the HEPs for the new, fire-       QU-C1           approach should be revised to treat the related human actions did not account for                       simplified approach for the non-fire actions the fire scenario-specific performance                           as a screening tool, and provide detailed shaping factors (PSFs), including time                           HRA for significant HEPs. Additionally, even available to complete action, etc. The formal                   the screening results should be reviewed for HRA of significant HEPs includes                                each fire scenario where the HEP is applied, referencing and considering procedures to                       including consideration of timing, lost perform actions, assessment of the impacts                       indications, spurious operations in the instrumentation needed for cues and                             scenario and other effects on the timing for execution and resulting impacts to the HPE,                      the HEP.
feasibility, timing of the event, performance shaping factors, evaluation of cognitive and execution error probabilities, scenario-specific equipment impacts that may affect the timing of the human interaction, as well as considerations of workload (for input to the dependency impact evaluation). Overall, it appears as if the Fire PRA treats the screening results of the method used for the non-fire human actions as detailed results.
However, as indicated in the diagram shown in the report, the method is a "simplified" Revision 0                                                                                                                                                       Page V-37

Florida Power & Light                                                                                                              Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR           Basis and Recommendation                   Disposition in Fire PRA Update method, and does not meet the requirements of a detailed HRA per the standard. Although the original HEPs in the internal events PRA included all of the relevant HRA factors, the fire-specific HEPs would have to consider the impact of fire on these factors in developing the HEP results.
The end result is an HRA that provides very general results, without specific application to a fire compartment or scenario, and a lack of detailed HRA for significant fire areas.
(This F&O originated from SR HRA-C1) 2-7            PTN system model changes in the fire PRA        PRM-B9         Systematic issue. System models were not    This F&O has been closed.
models are summarized in Tables D-1, D-2        SY-A 11        updated according to the SY-A and SY-B      The action taken to address this item was and D-3 of the Component and able                              SR requirements.                            specifically included in the focused-scope Selection Report 0493060006.001, Revision       SY-A14 Consider updating the system models and     Peer Review.
1, Appendix D, MULTIPLE SPURIOUS                SY-A15         their associated documentation according to OPERATIONS EXPERT PANEL EVIEW                  SY-A17          SY-A and SY-B SR requirements.
AND DISPOSITION OF OPEN ITEMS.
SY-A2 However, no additional documentation of the changes for PTN fire PRA has been              SY-A23 provided. Requirements under SY-A and          SY-A3 SY-B are not met as a result. In particular,    SY-A4 the requirements.of SY-A2, 3, 4, 6, 11, 14, 15, 17, and 23, as well as SY-B5 through        SY-A6 B15, do not appear to be met based on a         SY-B10 review of the documentation in the tables.      SY-Bi 1 Since the level of analysis and documentation for system models for Fire        SY-B12 PRA is expected to be similar to that           SY-B13 performed for internal events, the              SY-B14 documentation in the Appendix D tables          SY-B115 does not meet what is expected for this requirement. For example: on Table D-3,         SY-B5 page D-57, there appears to be a number of      SY-B6 changes to 'correct' logic. The original logic  SY-B7 is included in the system notebooks (e.g.,
DC power notebook). However, since this is      SY-B8 just corrected here, the internal events PRA    SY-B9 Page V.38 Revision 00                                                                                                                                                    Page V-38

Florida Power & Light                                                                                                                  Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                   Fire PRA SR             Basis and Recommendation                      Disposition in Fire PRA Update is not updated. Additionally, without proper development, it is impossible to determine if the change is accurate. In the DC power case, a reference to the power drawing is needed to ensure the logic is now correct.
On Table D-2, item 23, there are a lot of logic changes. However, there is no reason as to why the logic changes are needed or why the changes are made as written. In Tables D-2 and D-3, numerous references have been made to the comments /
recommendations from site engineers, which add insights but should not be considered as the sole modeling basis. (This F&O originated from SR PRM-B9) 3-10          The PTN Fire PRA used much lower HRR            FSS-C4          Transient fire evaluations conducted as         This F&O has been closed.
for the evaluation of transient fires. The HRR  FSS-D6          described in the Fire Scenario Report result    The action taken to address this item was for transient fires is based on the                             in screening fire damage to targets due to      specifically included in the focused-scope NUREG/CR-6850 HRR for electric motor            FSS-G1          the lower HRR which is believed to be non-      Peer Review.
fires (See F&O 3-10). The Fire Scenario                         conservative for developed fires involving Report (Report 0493060006.004) indicates                        ordinary combustible fuel packages such as that a transient NPP transient fire is better                   a trash can or trash bag.
represented by a temporary cable                                Use the NUREG/CR-6850 HRR for transient installation, which includes an ignition                        fires or provide alternate justification for an source. Based on this the Fire Scenario                         area-specific HRR based on the limiting fire Report indicates that the electric motor HRR                    that could occur within the area.
is used to describe transient fires. A review of the EPRI Fire DB of transient fires indicated the following: a) events in the DB indicated that the fire was either above 75 kw or could have been above 75 kw, if not suppressed. For example, one transient fire resulted in an automatic suppression system actuation, which was likely above 75 kw due to the sprinkler head being above 160 F as a result of the fire, b) a recent event at one of the peer review team members plant was above 75 kw. As a result of this review, and discussion amongst the peer review team Page V-39 Revision 0                                                                                                                                                         Page V-39

Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                      Fire PRA SR           Basis and Recommendation                    Disposition in Fire PRA Update members, the HRR for transient fires does not appear to be substantiated for the PTN FPRA. (This F&O originated from SR FSS-D6) 3-11          The screening criteria is defined in the          FSS-G2        NUREG 6850, Section 11.5.4.3 Step 3.c.1       This F&O has been closed.
Turkey Point Hot Gas Layer and Multi-                           recommends development of a conservative      The action taken to address this item was Compartment Analysis, (Report                                  HRR based on a combination of ignition        specifically included in the focused-scope H0493060006.006) methodology.                                  source and secondary combustibles that        Peer Review.
Compartments that don't screen are retained                     produce the highest HRR. This for further analysis. A concern identified with                 recommendation is provided to ensure that a the screening criteria involves the use of a                    conservative/realistic HRR is used to standard fire scenario for each analysis                        determine the potential for HGL formation.
rather than determining the most challenging                   Use of a non-conservative HRR may lead to fire scenario inherent to the analyzed                          underestimation of the potential for HGL compartment. This approach potentially                          formation and accordingly spread of masks the potential for forming an HGL in                       potentially damaging hot gas to adjacent the exposing compartment.                                       compartments.
(This F&O originated from SR FSS-G2)                           A realistic HRR should be developed for each fire compartment to ensure that the potential for formation of an HGL is appropriately assessed on a compartment by compartment basis.
3-13           Review of fire modeling in enclosed              FSS-C1        Systematic application of generic fire        This F&O has been closed.
compartments does not appear to accurately       FSS-G1        modeling results to define target damage      The action taken to address this item was consider the addition of HRR from                              when fire spread should be postulated.        specifically included in the focused-scope secondary combustibles. For instance a                         In situations where the generic fire modeling Peer Review.
switchgear cubicle fire located in fire zone                    treatments demonstrate fire spread to 71 is estimated to damage targets above the                     secondary combustibles, the scenario cubicle 4' horizontally and 7' vertically. The                  should assume full area damage. These switchgear cubicles are vented at the top so                    scenarios are candidates for detailed fire a fire in these cabinets can be expected to                    modeling that would be used to show fire spread to the cable trays above. The cable                      damage to PRA targets and the time to trays would be expected to ignite within the                    damage allowing credit for non-suppression plume of the cabinet which is estimated to                      probability.
be at least 3' wide. Given that an initial width of 3' is reasonable, fire spread and additional HRR due to the resulting cable tray fire only assumes 1' of fire spread along Revision 0                                                                                                                                                       Page V-40

Florida Power & Light                                                                                                                       Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR            Basis and Recommendation                           Disposition in Fire PRA Update the cable tray. Compared to the NUREG/CR-6850 guidance for flame spread along PVC cable (flame spread = 0.9 mm/sec) the estimation of HRR for these typical scenarios is nonconservative.
Realistic estimation of the scenario HRR is necessary to ensure the full impact of the fire on exposed targets is presented and that the effects of a damaging HGL may also be estimated.
(This F&O originated from SR FSS-C1) 3-14          According to the generic limitations            FSS-Cl          Systematic application of generic fire              This F&O has been closed.
contained in Attachment B of the Fire          FSS-G1          modeling results to define target damage            The action taken to address this item was Scenario Report (Report 0493060006.004)                        when fire spread should be postulated.              specifically included in the focused-scope the generic fire modeling treatments do not                     The generic treatments used in relatively           Peer Review.
account for the effects of hot gas layer                       small rooms should be scrutinized to ensure (HGL) on the correlations presented. The                       that any HGL interaction is considered and limitation indicates that because HGL is not                   accounted for if found to be significant.
considered that these correlations should not be used in enclosed areas with small volumes where a significant HGL thickness may form. Because this relationship is not considered plume temperatures may be underestimated because it is assumed that ambient temperature air is being entrained into the plume, resulting in cooler plume temperatures, rather than heated air from the hot gas layer. Entrainment of heated air into the fire plume results in higher damage heights because the plume remains hotter at higher elevations. (This F&O originated from SR FSS-C1) 3-2           Credit for fire compartment separation via      PP-B1           As noted in the description non-fire rated          This F&O has been resolved.
non-rated construction was commonly             PP-B2           construction is credited for separation of fire     The configuration and construction of non-noted, e.g., according to the FHA the walls                     compartments, however no Fire PRA                   fire rated barriers was confirmed using a of fire compartment 034 are not fire rated                      specific justification for the validity of the fire combination of information in the Fire and they provide separation from fire                           compartment is provided. This is considered         Hazards Analysis and supplemental plant compartments 036, 035, & 058. Separation                        a systematic issue for the FPRA.                    walkdowns. The anlaysis and related Revision 0                                                                                                                                                             Page V-41

Florida Power & Li-qht                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR              Basis and Recommendation                   Disposition in Fire PRA Update of FC 034 from the surrounding FCs is one                       Provide FPRA specific justification of the     documentation was updated to provide this of many examples where non-fire rated                           construction separating fire compartments,     information.
construction is credited for separation. Use                    where the barriers will substantially contain of this level of separation is acceptable                       the fire.
provided the separation is justified.
However, the justification does not appear to be provided for the FPRA. (This F&O originated from SR PP-B2) 3-3            A few cases of special separation are           PP-B1           Two instances were identified where spatial    This F&O has been resolved.
credited in the PB&P. Most notable are          PP-B3           separation is credited for the separation of   Openings between fire zones were separation of Fire Compartments 058 and                         fire compartments. No justification is         addressed with respect to targets on the 037 and 004 and 010. The FHA notes in the                       provided for this separation.                  other side of an opening which are within the write-up for fire zone 004: 'There is a partial                 Provide justification for the use of spatial   zone of influence of an ignition source.
height concrete wall on the South side of                       separation in the FPRA. If not justified,      Targets were evaluated for fire damage this room with a full height opening to Fire                    combine the compartments in the FPRA.          regardless of the zone in which they were Zone 10'. No justification is provided for this                                                                located.
separation, hence it is not clear that the                                                                     The multi-compartment analysis considered credited separation may be expected to                                                                         the mumpatent               anacent zons contain the effects of a fire. Accordingly the                                                                 the volume associated with adjacent zones effect of a fire beyond the identified fire                                                                    with openings between the zones in compartment boundary may occur. While                                                                          evaluating the potential for hot gas layer this effect would be expected to be identified                                                                 formation.
through performance of the multicompartment analysis the level of documentation provided in support of the PB&P does not satisfy the standard requirements. (This F&O originated from SR PP-B3) 3-4            The PTN self assessment points out that the     PP-B1           As discussed in the description                This F&O has been resolved.
FHA documents the use of active fire barrier    PP-B5           justification/discussion is not provided for   The walkdowns that were performed did not features as necessary for fire zone                             crediting active fire protection features in   observe any open fire doors (active separation. However in cases where fire                         barriers that are identified as non-fire rated features). The documentation for the fire compartment separation is provided by                           structures. It is not clear if active features scenario development process was updated unrated barriers there may be active                            such as fire dampers exist in these barrier    to provide the criteria and methodology that features that are not identified by the FHA                     segments because the FHA does not rely on      were used.
but credited by the Fire PRA. In such cases                     them for separation. Documentation should active fire barrier features may be                             be provided that clearly establishes what unknowingly credited for separation but not                     features are credited in such barrier b
Revision 0                                                                                                                                                         Page V-42

S Florida Power & Light                                                                                                                     Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR             Basis and Recommendation                     Disposition in Fire PRA Update adequately maintained by the fire protection                     segments and why makes them acceptable.
program. Because these elements were not                         Given the large number of barriers credited purposely identified within the development                       in the FPRA that are discussed in the FHA, of the Fire PRA it is unknown if the Fire                        but without discussion of active elements, Protection Program identifies all of the                         there are likely a number of undocumented necessary features. Because the Fire PRA                         active elements in these barriers.
does not formally define and justify these                        Determine the active fire barriers on barriers features this element is judged not met.                         credited in the FHA (not SSA), and provide (This F&O originated from SR PP-B5)                              justification for any active elements credited in the FPRA.
3-5           According to the Section 3.13 of the PTN          SF-Al           As discussed in the description no              This F&O has been resolved.
FPRA Summary Report the effect of an                             discussion was found that specifically          The low seismic spectra applicable to the earthquake on ignition source scenarios is                       addresses fire ignition source scenarios that   Turkey Point site have been validated via discussed in the IPEEE and Potential Fire                        may arise from an earthquake. Also, since       the IPEEE with respect to the potential for Related Vulnerabilities self assessment.                         these scenarios are not identified a            causing unique fire scenarios. Their Review of the Potential Fire Related                             qualitative assessment of their risk            potential for causing damage to pipes or Vulnerabilities self assessment did not                           significance is not included,                   tanks containing combustible gases or reveal an analysis that specifically                             The analysis provided in the Potential Fire     liquids or to initiation of electrical fires is addresses generation of fire ignition source                     Related Vulnerabilities self assessment         considered negligible.
scenarios which could result from an                             should be expanded to look for unique earthquake, nor does this assessment                             ignition source scenarios that may arise from address the potential risk significance of                       an earthquake and a discussion of the risk these scenarios. This assessment does                            significance of these scenarios should be identify fire vulnerabilities in terms of fuels,                 qualitatively assessed.
ignition sources, and oxidizers however these discussions are not specific to seismic events nor do they include evaluation of special ignition scenarios that may arise from an earthquake. (This F&O originated from SR SF-Al) 3-7           According to report PTN-PSA-7.01 The              IGN-Al          As discussed in the description the revised     This F&O has been resolved.
generic fire ignition frequencies provided in     IGN-B4          generic fire frequencies contained in FAQ       The guidance provided in FAQ 08-0048 NUREG/CR-6850 were used to establish the                         08-048 are not incorporated into the PTN        requires the use of the original fire ignition frequencies for PTN. While the                      fire frequencies nor is there justification for NUREG/CR_6850 fire frequency values as a use of these values is not entirely incorrect,                   their exclusion. This SR requires use of the    sensitivity study. Rather than perform two this SR requires the use of'current nuclear                       current nuclear power industry event history    analyses, the PTN analysis was developed power industry event history that includes                       or justification for data exclusion. Because    using those original values for the NFPA 805 power plants of similar type, characteristics,                   the fire ignition frequency methodology does Revision 0                                                                                                                                                               Page V-43

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR           Basis and Recommendation                      Disposition in Fire PRA Update and vintage.' Accordingly this requirement                      not address the data contained in FAQ 08-       application.
requires use of the EPRI revised generic fire                   048 this SR is considered not met. Use of       The application of the non-segregated bus frequency values included in FAQ 08-048 or                      the NUREG/CR-6850 values results in a           duct information from FAQ 07-0035 is not justification for its exclusion. Also, it appears                conservative estimate of CDF/LERF. FAQ          applicable as the plant does not use non-that FAQs 07-35 (bus ducts) and 08-44                           35 can have significant impact on fires in the  segregated bus duct. The connections to (MFW pump fires) were not incorporated into                     area of bus ducts. However, it is not           the station transformers are made using the FPRA. (This F&O originated from SR                          apparent if this is important for Turkey Point. cables. FAQ 08-0044 was also not needed IGN-A1)                                                         FAQ 44 can result in a lower MFW large fire     and the conservatism associated with frequency.                                      original method did not adversely affect the The fire ignition frequency information         results.
contained in FAQ 08-048 should be incorporated into the PTN fire ignition frequencies. Additional FAQs should also be incorporated into the FPRA.
3-8            Review of the plant-specific fire events for     IGN-A4         As discussed in the description review of the   This F&O has been resolved.
outlier experience indicates that some                          fires identified in Appendix A reveals fires events may have been considered outliers                        that may have become challenging had they       The scope of plant specific fire events were or unknown if the selection criteria had                        not been extinguished early. The selection      re-assessed with an expanded group of considered treatment of fires that are                          criteria for challenging fires contained in     plassesselwith panded fous of extinguished prior to full development as                       Appendix A is based on section C.3.3.1 of       plant personnel with particular focus on the potentially challenging. Several cases                          NUREG/CR 6850, however the criteria             subjective criteria from C.3.3.2. The results identified in Appendix A of the Fire Ignition                   contained in C.3.3.2 is not included; had the   of the re-assessment affirmed the previous Frequency Development Report, PTN-PSA-                          criteria of C.3.3.2 been included more fires    dispositions.
7.01 may have developed into challenging                        may have been selected as challenging or fires had they not been discovered and                          identified as unknown.
extinguished early in their development.
Fires 7, 8, 9, 21, 22, 27, 30, 31 appear to be                  The criteria for selecting challenging fires in potentially challenging fires (or unknown).                     Appendix A of the Fire Ignition Frequency See also the previous assessment from 9-                        Development Report, PTN-PSA-7.01 should
: 09. (This F&O originated from SR IGN-A4                         be revised to include the criteria contained in C.3.3.2 of CR/NUREG 6850 and the fire events should be revisited to determine if additional fires should be selected.
3-9            One situation was identified for which credit    FSS-C8         This finding is based on identification of      This F&O has been closed.
of a fire wrap is proposed. The FRANC            FSS-G1         credit for a wrap in the FRANC Excluded         The action taken to address this item was Excluded Events Table, Attachment E of the                      Events Table, Attachment E of the Fire          specifically included in the focused-scope Fire Scenario Report, (Report                                   Scenario Report, (Report 0493060006.004).       Peer Review.
0493060006.004) indicates that a fire wrap Revision 0                                                                                                                                                          Page V-44

0                                                                             0 Florida Power & Light                                                                                                                     Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                       Discussion                    Fire PRA SR             Basis and Recommendation                       Disposition in Fire PRA Update will be credited in Fire Zone 071. This fire                     Credit for the proposed fire wrap should be wrap protects an MCC 36 cable and is being                       addressed in the wrap integrity should be credited to exclude basic event 3806. No                         established with respect to fire resistance, technical basis for the fire resistance rating                   mechanical protection, and potential fire of this wrap was found in the FPRA nor is                        related exposure to which the wrap may be there justification for crediting this wrap                      exposed (direct flame impingement, HEAF, assuming mechanical damage, direct flame                         etc.).
impingement, or HEAF. Accordingly this SR is considered not met. During the walkdown, Thermo-lag was seen throughout the plant.
Thermo-lag has had problems in the past, and the rating would need to be justified prior to credit. (This F&O originated from SR FSS-C8) 4-10           The HRR used for fire modeling of the zone       FSS-C4          Severity factor is used extensively in the Fire  This F&O has been closed.
of influence is based on motor fires, which is   FSS-D3          PRA.                                             The action taken to address this item was substantially lower than the NUREG/CR            FSS-D5          Use the NUREG/CR-6850 HRR for transient          specifically included in the focused-scope 6850 recommended HRR of 317 kW. As a                             fires, or develop an accepted industry HRR       Peer Review.
result, the use of severity factor could         FSS-E2          approach (presently being discussed by potentially be double counting the lowered       FSS-G1          EPRI). Develop transient fire severity factors HRR for transient fires (note that even when                     based on the likely HRR and location of the severity factor is used as the                               overhead cables or location of equipment.
location/placement factor for transient fires,                   For example, if cable is 7 feet overhead, the it is dependent on the HRR in terms of the                       severity factor would be based on the zone of influence), if the severity factor                       minimum HRR that would damage the cable development is based on the NUREG/CR-                            at that distance. Additionally, the growth time 6850 HRR for transient fires. The severity                       can be used in determining non-suppression factor for transient fires discussed in Section                  time for generic cases, based on the latest 8.4 of the fire scenario report                                  FAQ 52. Finally, it is recommended that the (0493060006.004) does not provide                                current conditional probabilities in 7.1.2 table sufficient justification for a generic transient                 for electrical cabinets should not be used in fire severity factor. In addition, the severity                  the FPRA.
factor derived from an analysis of the number of fire events includes non-suppression results, and would therefore not be independent of any non-suppression probabilities applied later. FSS-C4 requires severity factors to be independent of other factors. It is noted, however, that the Revision 0                                                                                                                                                            Page V-45

Florida Power & Light                                                                                             Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR          Basis and Recommendation  Disposition in Fire PRA Update implementation of the severity factor in the PTN FRANC model did not involve the application of both severity factor and non-suppression factor in the same scenario.
Fire severity factor as discussed in Section 7.1.2 for low voltage electrical cabinets is not developed or applied consistently with the NUREG/CR-6850 methods. This is developed from a supplemental report (ERIN report, Supplemental Fire PRA Methods).
Additionally, fire propagation outside of the electrical cabinets is also dependent on the non-suppression probability. Therefore, some dependency exists in this data if used in conjunction with a non-suppression factor.
The numbers listed in 7.1.2 for electrical cabinets were derived using the total number of cabinet fires in the denominator, rather than the number of fires of the specific panel type. Due to this incorrect derivation of the conditional probabilities for fire propagation outside of the cabinets, the conditional probabilities thus developed (and applied in the FRANC model for low voltage cabinets) could potentially be low by an order of magnitude (non-conservative). For both the transient fires and low voltage cabinet fires, the severity factors are basically developed using fire events data from the EPRI report. Given the fire data duration and damage is a result of multiple factors (growth, suppression, severity, location, etc), and given the fire data often does not have sufficient information to make a reasonable determination of either the fire size or whether a fire propagated outside the cabinet, the peer review team determined that the use of fire events data for developing the above severity factors is not appropriate.
Revision 0                                                                                                                                 Page V-46

Florida Power & Light                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update (This F&O originated from SR FSS-C4) 4-11           The argument in Section 8.1 of the fire          FSS-C4          The location/severity factor applied should   This F&O has been closed.
scenario report about the application of the     FSS-D3          have already accounted for the probability    The action taken to address this item was probability of an ignition source being          FSS-D5          that the target is located within the impact  specifically included in the focused-scope located within an area around the target may                     area around the transient fuel and an         Peer Review.
not be correct. This is because the              FSS-G1          ignition source is located within an area in application of a location factor for the         FSS-H6          the vicinity to ignite the transient fuel.
transient fuel to the fire ignition frequency                    Revise Section 8.1 and do not apply has already accounted for the probability of                     separate location factors for transient fuel the target being within the influence zone of                     and ignition source simultaneously in any a fire. As such, ignition is a given condition,                  fire scenario.
With the apportioned frequency, the target must be located within the impact area around the transient fuel and an ignition source must be located within an area in the vicinity to ignite the transient fuel. (This F&O originated from SR FSS-C4) 4-12           Except for the MCR fire scenarios, no            FSS-E3          Mean values and uncertainty intervals for     This F&O has been closed.
scenario-specific fire modeling has been          FSS-H9          the parameters used for modeling the          The action taken to address this item was performed to provide a mean value of, and                         significant fire scenarios have not been      specifically included in the focused-scope statistical representation of, the uncertainty    UNC-A2          provided.                                     Peer Review.
intervals for the parameters used for                             Consider developing mean values and modeling the compartments with significant                        uncertainty intervals for the parameters used fire risk contributions. Therefore, for                           for modeling the significant fire scenarios.
compartments other than MCR, only the results of conservative, generic fire modeling developed in the Generic Fire Modeling Treatments report were applied to the analysis of fire scenarios. (This F&O originated from SR FSS-E3) 4-13          Uncertainties associated with cases where         FSS-E4          The required uncertainty has not been         This F&O has been closed.
cable routing has been assumed (e.g., the         FSS-H9          evaluated / performed.                        The action taken to address this item was EXCLUDEDEVENTS table has assumed                  UNC-A2          Consider investigating the uncertainties      specifically included in the focused-scope that certain cables are not routed through                        associated with cases with assumed cable      Peer Review.
selected areas based on walkdown or                               routing.
engineering judgment) have not been investigated with a documented basis. (This Revision 0                                                                                                                                                         Page V-47

Florida Power & Light                                                                                                                   Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR           Basis and Recommendation                     Disposition in Fire PRA Update F&O originated from SR FSS-E4) 4-14           The Hot Gas Layer and Multi-Compartment           FSS-G3        Multi-compartment fire scenarios' risk          This F&O has been closed.
Analysis report (H0493060006.006,                 FSS-G6        significance is not yet evaluated.              The action taken to address this item was Revision 0) performed a screening                               Complete the detailed multi-compartment         specifically included in the focused-scope evaluation of the need for hot gas layer and                    analysis and add the discussion on multi-       Peer Review.
multi-compartment analysis, and identified                      compartment fire scenarios' risk contribution.
scenarios/zones that warrant further evaluations. No detailed multi-compartment analysis is completed (still in progress) in this report, and no discussion on multi-compartment fire scenarios' risk contribution is provided. (This F&O originated from SR FSS-G6) 4-15           Treatment for transient fire damage to            FSS-D6        Significant modeling issues resulting in        This F&O has been closed.
targets is measured from the compartment          FSS-G1        many transient fire scenarios being             The action taken to address this item was floor rather than the height of the transient     FSS-H6        screened during detailed scenario analysis. specifically included in the focused-scope fuel package that is typically considered.                      The results of the FPRA are therefore           Peer Review.
PTN FSS report section 8.2 states, "Cable                       potentially non-conservative for the analyzed trays (or the lowest tray within a stack of                     detailed scenarios. Transient fire evaluations trays) that were at least 5.8' off the floor                    conducted as described in the Fire Scenario were considered beyond the zone of                              Report result in screening fire damage to influence of the transient fire for nonqualified                targets that are located > 5.8' above the floor cables." This apparently was based on the                       which is believed to be non-conservative for lowered HRR values (See F&O 3-10) used                          developed fires involving ordinary for the transient fires, as well as the transient               combustible fuel packages such as a trash fire being located at the floor. This may                       can or trash bag. In response to this concern result in the reperformance of transient fire                   it was pointed out that the thermal plume walkdown if the transient fire HRR values                       component relies on empirical relationships need to be updated. Discussion with FP&L                        between the source strength and the following the onsite review provided some                       distance between the virtual origin of the fire basis for the damage height (indicating that                    and the target. The fire plume begins to transient fires above the floor will have an                    entrain air at the lowest point of burning, overall lower average surface HRR).                             which defines the base of the fire; normally However, the supplemental discussion was                        at the floor. However this argument ignores still considered inconsistent with past events                  the potential that a fire could begin burning and existing guidance on analysis of                            at the top of a fuel package thus elevating its transient fires. (This F&O originated from SR                   base. At a minimum, during the initial period FSS-H6)                                                         of burning, damage temperatures generated Revision 0                                                                                                                                                          Page V-48

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update by the fire would likewise be elevated. Over time the base of the fire may change due to collapse of the fuel package or burning away of the fuel, however the empirical model presented did not develop these ideas as a reason for assuming that the base of the fire is at the floor for its entire duration.
The transient fires should be considered to be above the floor level in the analysis.
4-17           Per summary report, Task 9 is fulfilled with    CF-B1          Incomplete evaluation and document for           This F&O has been resolved.
the NISYS SSD database, "PTN NFPA 805                           circuit failure. The NISYS DB can include        Circuit failure probability was considered for Database.mdb". This database has been                          identification of when spurious operation        high risk scenarios and only in cases where significantly expanded for the NFPA 805                        may occur, but does not provide the circuit      doing so would result in a reduction in total tasks. A sample circuit analysis worksheet                      analysis or circuit failure probability analysis risk. Additional details with respect to circuit (e.g., for component 20ASB/G3) has                             needed to support the FPRA.                      configuration and raceway type have been signatures at the bottom, which were not                        Provide a documented basis, and detailed         added to the altered events table.
populated yet. The NISYS circuit analysis is                   circuit analysis for any spurious operation an Appendix R type circuit analysis and                         probability used in the FPRA per Tasks 9 does not identify the circuit failure modes                     and 10 of NUREG/CR-6850 (or equivalent).
and address likelihood of failure. Failures of the required cables identified are assumed to have a probability of 1.0 unless specifically modified in the ALTEREDEVENTS table of the FRANC model. The treatment of the circuit analysis seems to be bounding (i.e., the likelihood was not part of the analysis). Although Appendix D of the fire scenario report states the bases for the altered FRANC event probabilities, it seems that there is no linking between the altered probabilities and the circuit analysis package. The majority of the altered events are based on operator manual actions while some based on the simple spurious actuation probabilities from NUREG/CR-6850, which were based on specific evaluation (with no basis provided in the FRANC database), but do not directly linked to any specific circuit analysis Revision 0                                                                                                                                                           Page V-49

Florida Power & Liaht                                                                                                                     Attachment V - Fire PRA Qualitv Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR            Basis and Recommendation                      Disposition in Fire PRA Update worksheet. Since the "basis" column of the Altered event table in the FSS report does not appear to include sufficient documentation to allow review/peer review of the results and the NISYS database does not include the analysis, the analysis (not the results) has not been documented. The evaluation and documentation of the review of the fire-induced circuit failure modes and the assignment of the appropriate industry-wide generic values to their conditional failure probabilities for risk-significant contributors based on the specific circuit configuration under consideration should be included in the circuit failure report for Tasks 9 and 10. (This F&O originated from SR CF-B1) 4-19           The new fire-specific safe shutdown actions       HRA-B2          Risk significant human actions should be        This F&O has been closed.
identified and incorporated into the PTN Fire     HRA-B3          defined in accordance with SR HR-F2.            The action taken to address this item was PRA have not been defined (even for the                           Identify the risk-significant new fire-specific specifically included in the focused-scope HR-F2 risk-significant actions) by specifying (a)                       safe shutdown actions and define these          Peer Review.
accident sequence specific timing of cues,                        actions in accordance with SR HR-F2 and and time window for successful completion                         HRA-B3.
(b) accident sequence specific procedural guidance (e.g., AOPs, and EOPs) (c) the availability of cues and other indications for detection and evaluation errors (d) the specific high level tasks (e.g., train level) required to achieve the goal of the response, or the complexity of the response. (This F&O originated from SR HR-F2) 4-23           The time available and time required to           HRA-C1          Systematic issue. The evaluation of the time    This F&O has been closed.
complete actions were not evaluated for the       HR-G5           available and time required to complete the     The action taken to address this item was new, risk significant fire-related safe           SF-A5           risk-significant actions is required.           specifically included in the focused-scope shutdown actions. PTN FPRA HRA report                             Evaluate time available to complete the risk-   Peer Review.
Tables A-1 and A-2 include evaluation of the                      significant fire-related safe shutdown time available to complete actions. However,                      actions.
the point in time at which operators are Revision 0                                                                                                                                                            Page V-50 0

0 Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                   Fire PRA SR              Basis and Recommendation                   Disposition in Fire PRA Update expected to receive relevant indications are not evaluated. (This F&O originated from SR HR-G4) 4-24           It appears that the reasonableness of risk-    FQ-C1           The reasonableness of risk-significant, post- This F&O has been closed.
significant, post-initiator HEPs relative to   HRA-C1           initiator HEPs relative to each other should The action taken to address this item was each other was not yet reviewed in the                         be reviewed and checked in the scenario       specifically included in the focused-scope scenario context, plant history, procedures,   HR-G6           context, plant history, procedures,           Peer Review.
operational practices, and experience. (This   QU-C1           operational practices, and experience.
F&O originated from SR HR-G6)                                   Review the reasonableness of risk-significant, post-initiator HEPs relative to each other in the scenario context, plant history, procedures, operational practices, and experience.
4-25           Uncertainty characterization of the HEPs       HRA-C1          HEP uncertainty characterization is needed    This F&O has been closed.
developed for the Fire PRA was not             HR-G8           for the evaluation of uncertainty in the      The action taken to address this item was (This  F&O originated from SR HR-G8) provided.                                      QU-E4           overallllyincriskin results.-cop                    specifically included in the focused-scope UNC-A1          Develop uncertainty characterization of the   Peer Review.
HEPs used in the Fire PRA (especially for those risk-significant HFEs).
4-4            There does not appear to be a document         PRM-B2          SR PRM-B2 and NEI 07-12 require proper        This F&O has been closed.
recording the proper resolution of                             disposition of the deficiencies from the      The action taken to address this item was deficiencies from the previous peer review of                  previous peer review that may adversely       specifically included in the focused-scope the internal events model. In addition, a peer                 affect the accuracy of the Fire PRA model. Peer Review.
review or gap assessment of the major                          CDF has been significantly reduced (more changes since the previous peer review of                      than an order of magnitude) since the last the internal events model does not appear to                    internal events peer review. Additionally, exist (a gap database was provided, but with                   methodology changes have occurred, no supporting documentation). Finally, a gap                   including use of the HRA calculator and use assessment of the PRA standard changes                         of a new CCF model.
from RG 1.200 Rev. 1 to Rev. 2 does not                         Resolve all significant deficiencies from the appear to exist. (This F&O originated from                     previous peer review that may adversely SR PRM-B2)                                                     affect the accuracy of the Fire PRA model results, if not already completed, and document these dispositions. Provide a gap assessment or new peer review on the internal events PRA, latest revision.
Revision 0                                                                                                                                                       Page V-51

Florida Power & Light                                                                                                                    Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update 4-5             Documentation of the changes in the             PRM-C1          The level of documentation provided in          This F&O has been closed.
Internal Events PRA model to develop the                        Section 4 and Appendix D of the Fire PRA        The action taken to address this item was Fire PRA model are primarily provided in                        Component and Cable Selection Report            specifically included in the focused-scope Section 4 and Appendix D of the Fire PRA                         (Report 0493060006.001) does not meet the       Peer Review.
Component and Cable Selection Report                             SRs for IE-D, AS-C and SY-C.
(Report 0493060006.001). However, no                             Include additional documentation to the level additional, detailed documentation of the                        satisfying the SRs for IE-D, AS-C and SY-C.
changes is provided. Based on a review of the documentation in the tables in these sections (e.g., Tables 4.1-2, D-1, D-2, D-3, etc.), SRs for IE-D, AS-C and SY-C, etc. are not met. Since the level of analysis and documentation for Fire PRA model is expected to be similar to that for internal events, the documentation in the above sections and tables does not meet what is expected for this requirement. Let's look at some examples: on Table D-3, page D-57, there appears to be a number of changes to
                'correct' logic. The original logic is included in the system notebooks (e.g., DC power notebook). However, since this is just correction here, the internal events PRA is not updated. Additionally, without proper development, it is difficult to determine if the change is accurate. (This F&O originated from SR PRM-Cl) 4-6            The Fire Scenario Report (0493060006.004)        FSS-A2          Circuit failure modes are necessary for the     This F&O has been closed.
Appendices A, D, and E, and the FRANC                             assessment of circuit failure (e.g., hot short) The action taken to address this item was model document the equipment failure                             probabilities and required for meeting SR       specifically included in the focused-scope modes for each fire scenario. However,                           FSS-A2. The method used in the PTN FPRA         Peer Review.
circuit failure modes associated with failures                   can significantly over estimate the likelihood of the required cables were not identified or                    of the circuits causing such particular failure documented. Relevant circuit failure modes                       modes as spurious operation.
are necessary for the assessment of circuit                      Document circuit failure modes for the failure (e.g., hot short) probabilities. For                     required cables for each fire scenario most components, there is no differentiation                     evaluated.
in the FPRA between failure modes that can result due to failure of each cable/circuit. As Revision 0                                                                                                                                                           Page V-52 0

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR           Basis and Recommendation                       Disposition in Fire PRA Update a result, the fire scenarios assume each failure mode would occur from damage to all cables identified in the SSA. In order to refine the fire scenario under CF tasks, the circuit failure would need to be provided for each risk relevant circuit. (This F&O originated from SR FSS-A2) 4-8            The process of defining fire scenarios with      FSS-A5         Issue with potentially significant impact.      This F&O has been closed.
the source/target combinations and its                          For risk-significant fire compartments,         The action taken to address this item was FRANC implementation process are such                          develop additional fire scenarios such that      specifically included in the focused-scope that the risk contribution of each risk-                       specific targets are determined based on the     Peer Review.
relevant ignition source was characterized,                    location of each target (e.g., affected tray)
The evaluation and results are documented                      relative to the ignition sources in the room.
in the fire scenario report and the FRANC model. However, the analysis does not appear to always differentiate between targets (e.g., cables in different cable trays).
For a number of the top 10 scenarios (i.e.,
full zone burn-out scenarios), it appears all fires damage all equipment and all cables in all trays (without specific knowledge of where each of the targets are located, for example, in the trays). Basically, the process for developing detailed scenarios for all significant fire compartments has not been completed in the FPRA for CDF and LERF.
For example, see scenarios in 096, 019 and 020 (base case CDF scenarios). (This F&O originated from SR FSS-A5) 4-9            It appears that the application of severity      FSS-C4        This appears to be a systematic issue.           This F&O has been closed.
factors is inconsistent between the FRANC        FSS-G1        However, this may also be just a                 The action taken to address this item was model and those listed in Appendix A of the                    configuration control and/or documentation       specifically included in the focused-scope Fire Scenario Report (0493060006.004). In                      issue. Severity factor should be applied to all  Peer Review.
the FRANC model, severity factor is only                        applicable scenarios to derive realistic result.
used for transient fires and as the split                      Reconcile the differences between the fraction between severe and non-severe                          FRANC model and the fire scenario report.
MCC fires. For transient fires, it appears that                Apply the severity factor or non-suppression the severity factor is used in the sense of a                   factor to all applicable scenarios in a manner Revision 0                                                                                                                                                          Page V-53

Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                      Fire PRA SR          Basis and Recommendation                   Disposition in Fire PRA Update location factor associated with the                             consistent with the methodology and data placement of the transient fuel. In Appendix                    discussed in the fire scenario report.
A of the fire scenario report, severity factor is also used for oil fire, pump fire, and electrical cabinet fire, in addition to transient fires and MCC fires. Further, the values of the severity factor used in the FRANC model and Appendix A of the fire scenario report do not match. The bases for neither were documented for each individual scenario.
Also, the severity factor values used for the oil and pump fires do not appear to be consistent with the tabulated values given in Table 5-1 of the fire scenario report. It appears that severity factors or non-suppression factors can be applied to many more detailed scenarios in the FRANC model to make the estimate of the detailed scenario risk more realistic. In addition, the scenario ignition frequencies listed in Appendix A do not appear to be consistent with those used in the FRANC model. (This F&O originated from SR FSS-C4) 5-11           Review of Turkey Point NISYS NFPA 805             ES-B1         The deviation between the ESFAS             This F&O has been resolved.
Compliance Assessment Database within                           components actuation and Control Room SI    The circuit analysis process used for the the Cable Routing and Respective                                components actuation should be disposition  project has been confirmed to be consistent Equipment table, it was noticed that the                        and reconcile, to ensure Fire Safe Shutdown with the latest industry guidance (NEI 00-Spurious ESFAS signal                                           / Appendix R equipment are appropriately    01). In addition, the asymmetry was "Spurious/ESFASILacks/Analysis" have total                      credited in the Fire PRA.                   discussed with plant staff and confirmed to of 56 respective components impacted. Unit                      Reconcile the FPRA component list with the  be reflective of the actual plant design and 3 Train A SI signal from the Control Room                       SSA component list for equipment impacted   configuration.
                "3MRASI/3CO6/3QR431006" have total of 29                        by an SI signal respective components impacted, Unit 3 Train B SI signal from the Control Room "3MRBSII3CO613QR45/006" have total of 28 respective components impacted, Unit 4 also have similar components impacted. The concern is the potential mismatch between FPRA and the SSA component lists. (This Revision 0                                                                                                                                                       Page V-54

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                   Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update F&O originated from SR ES-B1) 5-13          Turkey Point FPRA Summary Report               FQ-A3            It appears that there is inconsistent basic  This F&O has been resolved.
NUREG/CR-6850 Task 16 Report No.                              event mapping between the database files. The identified data differences were 049306006.005 Rev. 1 Tables A-1, A-2, B-1                      A sensitivity run was performed by copying    reviewed and confirmed to be reflective of and B-2 documented the Units 3 &4 Fire                         the U4 events to the U3 tables, and re-       the design and layout of the units.
PRA quantification Results for both CDF and                   evaluated U3 CDF. The results are the top     Additional comparison of the quantification LERF for all fire scenarios that were                         scenario in 96 dropped from 4.5E-05 to 1E-    results between the two units was also quantified. Scenario 096-A was randomly                        06. Based on this, the error appears to be    performed to ensure that any significant picked review for both Units 3 & 4. The                        significant.                                  differences in results are consistent with the CDF/LERF results are consistent between                        Need to ensure that the altered events table  actual unit differences. Various the Summary Report and Zone Scenarios in                       is correctly developed for both U3 and U4 for asymmetries in the plant layout were database files, Unit 3 CDF                                     the CDF and LERF quantification.              identified.
               "PTNFIRE W LERF MH ESF.mdb", Unit 3 LERF "PTNFIREW_LERFMHESF.mdb",
Unit 4 CDF "U4PTNFIRE W LERFMHESF.mdb",
and Unit 4 LERF U4PTNFIRE W LERF MH ESF.mdb".
However, reviewing the Altered Events table in each database files shows inconsistent basic events impacted between Unit 3 and
: 4. Unit 3 have no basic event impacted, while Unit 4 have 9 basic events listed. (This F&O originated from SR FQ-A3) 5-16          Review of PTN Tasks 8 and 11 Report            FQ-A3           The discussion between the two reports are    This F&O has been closed.
0493060006.004, Rev. 1. Page 23, Section       FSS-Al          inconsistent regarding to misc hydrogen fire. The action taken to address this item was 7.5.2, states "no hydrogen fires other than                    Incorrect apportioning the fire frequency and specifically included in the focused-scope turbine/generator have been postulated. The                    define appropriate fire scenarios could have  Peer Review.
basis appears to be that they use excess                       significant impact to the CDF and LERF flow check valves to limit H2 release.                         results.
Question was asked during the review, the                      Address Miscellaneous H2 fires in the response said "The small quantify of                           FPRA, identifying other compartments hydrogen downstream of the check valves                        containing hydrogen piping.
and its potential leakage will result in small accumulations of hydrogen and are unlikely to result in combustible concentrations of hydrogen in any area of the plant". However, further review of PTN NFPA 805 Fire Ignition Revision 0                                                                                                                                                      Page V-55

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                        Discussion                       Fire PRA SR            Basis and Recommendation                   Disposition in Fire PRA Update Frequency Report PTN-PSA-7.01, Rev. 2, Fire Compartment Ignition Source Data Sheet (ISDS) for compartment 082 "Unit 4 Auxiliary Transformer Area" indicate that BIN 19 - Misc. Hydrogen Fires was identified in this area. For example, H2 feed to the VCT should be looked at. (This F&O originated from SR FSS-A1) 5-3            Based on a review of the Turkey Point FPRA           ES-D1          The report identifies multiple cases with    This F&O has been closed.
Component and Cable Selection Report                                confirm or investigation required. Some of 049306006.001 Rev. 1, the documentation is           PRM-t3c                re ionvstirequired.                  The action taken to address this item was sufficient to support the supporting requirements. PRM-B4         the resolutions were documented as           specifically included innthe focused-scope The documentation issues identified inthe F&Os                       addressed, but there some cases with no      Peer Review.
generally include suggestions to clarify                            documentation of resolution or justification information in the calculation to accurately reflect                can be found. The PRA documentation does the process followed inthe PRA. ES-D1-01                             not clearly show the resolution of the MSO concerns a finding related to the MSO attachment,                   items as required by the standard. This is This attachment needs to be updated to clearly                       considered a documentation concern document how each MSO was addressed in the                           because the resolution of these items can model in accordance with the actions from the                       be found in the PRA model. Suggestion to MSO expert panel. Another example: scenario 45                      update Appendix A with a clear resolution of involving (diesel overload) has a note to complete the on-going evaluation; scenario 34 (loss of SG                     all MSO identified open items.
inventory) has a note to verify total flow rates;                   Provide documentation on open items from ensure adequate Fire PRA documentation.                              the MSO expert panel. Include modeling in Another example is scenario 7, 'Normal letdown                       the FPRA of any known scenarios that are fails to isolate and inventory is lost to the                        found to be an issue.
pressurizer relief tank (PRT)', which has an open action to revisit PRA model structure for letdown isolation. The FPRA documentation doesn't discuss the disposition for this scenario and the FPRA doesn't model it. (This F&O originated from SR ES-D1) 5-5            Turkey Point FPRA Human Failure                      ES-A2          Any alarm procedure that does not require    This F&O has been closed.
Evaluation Report 0493060006.002, page 6             ES-C2          alternate confirmation will need to be       The action taken to address this item was states that a simulator review was performed         HRA-A3         reviewed and either dispositioned or         specifically included in the focused-scope to identify instrumentation that should be                          assessed for equipment to be added to the    Peer Review.
explicitly modeled in the FPRA, including            HRA-B4         FPRA component list. There does not "Identify any alarms or indications that would                      appear to be an extensive review of these lead operators to take immediate control                            alarm response procedures for either actions without further verification."                              identification of instrumentation failures Revision 0                                                                                                                                                           Page V-56

Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update Appendix C of the same report, page C-3                         leading to a trip or causing an operator to has a section asking which annunciator tiles                   shutdown plant equipment.
cause an operator to take immediate action.                    Review alarm response procedures to The response is "Operators will confirm the                    address this issue and document signal with an alternate indicator before                      appropriately. This aligns with the guidance taking any action." This answer is not                         in section 2.5.5 of NUREG/CR-6850.
specific, and may not always be procedurally correct. In a typical NPP, each annunciator sends the operators to some alarm response procedure. The alarm response procedure will typically require confirmation using an alternate indicator; however, this is not always the case. (This F&O originated from SR ES-A2) 6-1            No documentation is provided of a               ES-A3          Systemic issue                                 This F&O has been closed.
comprehensive review of fire impacts to         ES-A4          Perform a review of fire scenario equipment    The action taken to address this item was plant equipment for unique initiating events. ES-DI          impacts to identify fire-specific initiating   specifically included in the focused-scope The MSO list includes combinations of                           events. Examine groups components that         Peer Review.
spurious operation components. However,          PRM-B3         can be disabled by a single fire and include this review does not include consideration of    PRM-B4         the potential for a single spurious event.
combinations of fire-induced failures that can                  Examine these equipment impacts in terms lead to an initiating event. Additionally, there                of plant response, timing, success criteria, did not appear to be a review of screened                       and the effects on the operability and initiating events from the internal events                     performance of operators and mitigating PRA, other than ISLOCA pathways. (This                         systems. For each scenario, identify a new F&O originated from SR ES-A3)                                  fire-specific initiating event if no existing initiating event bounds or adequately represents the equipment impacts.
6-10          The altered events table in the FSS report       FQ-Al          This approach is not consistent with the level This F&O has been resolved.
includes several instances where a single        FQ-A4          of detail modeled elsewhere in the PRA.        The methodology and the analysis has been basic event combines a hot short spurious        HRA-E1         Translate specific failure modes into basic    updated to eliminate the use of this operation likelihood with an HEP to recover                     events and avoid combining disparate failure   approach. The use of altered events for the spurious operation. For example,             HR-I1          modes into combined basic events,              spurious probability is used only as required ORZR30455C represents a combination of           HR-12                                                         and a singular value.
spurious opening of a PORV and operator          QU-A3 human error probability to close the PORV.
Supporting requirement FQ-Al addresses the need to translate specific failure modes Revision 0                                                                                                                                                        Page V-57

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update into basic events. Embedding an HEP with a spurious operation likelihood bypasses this requirement, and this approach is not consistent with the level of detail modeled elsewhere in the PRA. Also, the approach prevents the ability to address the state of knowledge correlation. (This F&O originated from SR FQ-Al) 6-11          The FPRA models through the Altered             HRA-A1          The operator recovery actions were based      This F&O has been closed.
events table several 0.1 values for recovery    HRA-A2          on proposed new procedures instead of the     The action taken to address this item was actions that are not in the safe shutdown      HRCl            existing ones, don't reflect the as built as  specifically included in the focused-scope analysis and are not in the fire safe                          operated plant and are not confirmed to be    Peer Review.
shutdown procedures. There are about 198        HRA-D2          feasible.
unique instances of such recovery actions. HRA-El          Ensure that all FPRA human failure events For example, event AHFPTRNAMAN, in             HR-El           reflect the as-built, as-operated plant, and area 63, is not included in operations                          that they are proceduralized. Verify all procedure 0-ONOP-01 6.10. This human            HR-H2           credited actions, including those modeled in interaction modeling doesn't reflect the as    HR-I1           the HRA and those included in the altered built as operated plant, and no evaluation of   HR-12           events report, are included in the plant feasibility is documented for these actions.                    operational procedures.
Discussions with the FPRA development team indicated that the post-fire operating procedures will be updated to incorporate the new recovery actions and feasibility will be evaluated at that time. The FPRA will need to be updated, as necessary to reflect the outcome of feasibility evaluations. (This F&O originated from SR HR-El) 6-12           Based on a review of the FPRA HRA              HRA-Al          Step not performed                            This F&O has been closed.
Report, no talk-throughs or reviews appear     HRA-A4          Perform talk-throughs or reviews of           The action taken to address this item was to have been made with plant operations         HRA-E1          procedures and sequences of events with       specifically included in the focused-scope and training personnel of procedures and                       plant operations and training personnel to    Peer Review.
sequences of events to confirm that             HR-E3           confirm that interpretation of the procedures interpretation of the procedures by FPRA       HR-I1           is consistent with plant observations and developers is consistent with plant             HR-12           training.
observations and training. The report indicates that a simulator review was Revision 0                                                                                                                                                       Page V-58

0 Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update performed to identify instrumentation that should be explicitly modeled in the fire PRA as generally required to shutdown the plant or to perform credited operator actions, and a review agenda is provided in Appendix C, but no documentation of such a review is provided. (This F&O originated from SR HR-E3) 6-13           No simulator observations or talk-throughs      HRA-A4          Step not performed                            This F&O has been closed.
with  operators  have been   performed confirm   the response  models   for fire to     HRA-E1          Perform simulator observations or talk-       The action taken to address this item was scenarios modeled. (This F&O originated          HR-E4           throughs with operators to confirm the        specifically included in the focused-scope from SR HR-E4)                                   HR-Il           response models for fire scenarios modeled. Peer Review.
HR-12 6-15          The documentation of credited recoveries in      HRA-E1          Systematic Issue                              This F&O has been closed.
the altered events table in many instances is    HR-I1           Perform analyses to verify operator vague (e.g., "Manual action to restore bus,"     HR-12           recoveries and repairs credited in the        The action taken to address this item was and "failure probability of new inside control                   Altered events Table in Appendix D are        specifically included in the focused-scope room HEP, required in less than 60 minutes       PRMB9           feasible, prior to crediting any recovery via Peer Review.
(3AA, DC 205)"). The effort to evaluate                          manual operation of the equipment.
these actions for inclusion in the SSA and feasibility would be facilitated by more detailed descriptions of the actions. PTN system model changes in the fire PRA models are summarized in Tables D-1, D-2 and D-3 of the Component and Cable Selection Report 0493060006.001, Revision
: 1. However, no additional documentation of the changes is provided. Requirements under SY-A and SY-B are not met. Repair of components that are spuriously operated or fire damaged is modeled using the altered events table of Attachment D, FRANC Altered Events Table, of the Fire Scenario Analysis Report 0493060006.004, Rev 1.
However, the substitution does not include a verification that the actions are possible or feasible. (This F&O originated from SR Revision 0                                                                                                                                                        Page V-59

Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR          Basis and Recommendation                      Disposition in Fire PRA Update HRA-E1) 6-16          The HEP dependency evaluation produced            FQ-Cl         Unreasonably low values assigned to            This F&O has been closed.
several dependent HEPs on the order of 1E-        HRA-C1        dependent HEP combinations.                    The action taken to address this item was 11 to 1 E-1 3 (for example, cases 62, 93, 110    QU-C2         Assign a floor for dependent HEP               specifically included in the focused-scope and 96). NUREG-1792, "Good Practices for                       combinations using a justified minimum         Peer Review.
Implementing Human Reliability Analysis,"                      value.
recommends that the total combined probability of all the HFEs in the same accident sequence/cut set should not be less than a justified value. NUREG-1 792 suggests that the value not be below -1 E-5, since it is typically hard to defend that other dependent failure modes that are not readily anticipated cannot occur. However, some industry PRAs are using a floor value of
               -1 E-06. (This F&O originated from SR HRA-Cl) 6-20          The parametric uncertainty associated with        CF-A2         Step not performed                             This F&O has been resolved.,
conditional circuit failure probabilities are not UNC-A2        Step not performed                             Parametric uncertainty has been performed evaluated and are not incorporated into the                                                                    for CDF and LERF for each unit's FPRA.
model. (This F&O originated from SR CF-A2) 6-3            For the FPRA, no accident sequences were         AS-Al         Required step not performed.                   This F&O has been closed.
identified beyond those modeled by the           AS-A9         Perform a review of FPRA scenarios to          The action taken to address this item was internal events PRA. The FPRA accident           LE-Al         ensure that the existing event tree structures specifically included in the focused-scope sequence accident progression, success                          accurately model the specific FPRA initiating  Peer Review.
criteria and timing are therefore based on        LE-A2         events, including considerations of timing, the internal events PRA. Consideration            PRM-B14       plant response, and human interactions.
should be given, however, to success              PRM-B15 criteria and timing specific to the FPRA. For example, no evaluation is made of the timing      PRM-B5 associated with RWST draindown. Also,             PRM-B7 RWST draindown may require sump recirculation, which is not presently represented in the non-LOCA transient event tree accident sequences. This step has not been performed and finding is made Revision 0                                                                                                                                                        Page V-60

Florida Power & Light                                                                                                                          Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                        Discussion                         Fire PRA SR           Basis and Recommendation                       Disposition in Fire PRA Update to include such considerations in the FPRA development and documentation. Also, when any new FPRA initiating events are identified as part of resolving ES F&Os, consideration will need to be made of the accident sequence accident progression and timing associated with any new accident sequences. (This F&O originated from SR PRM-B7) 6-4            No FPRA modeling appears to have been                   AS-Al         Step not performed. Potentially significant      This F&O has been closed.
made to address the actions directed by the              AS-Al0        impact on FPRA accident sequences and            The action taken to address this item was fire safe from    the shutdown     procedures actions directed    by thethat deviate EOPs.    (This AS-A4         results.specifically results.icldd                                                included in the focused-scope nth    ocsd-cp Modeling of the following procedural             Peer Review.
F&O originated from SR PRM-B6)                           AS-A5         responses to a fire may be needed: 1)
HRA-A1        equipment is disabled to preclude spurious HRA-A2        actuations; 2) human actions to isolate HR-El         unprotected equipment; 3) human actions to manually operate protected equipment.
Some of the above human actions could PRM-B5        also induce new sequences not traditionally PRM-B6        covered in the Internal Events PRA. New sequences to account for these effects may also need to be incorporated into the Fire PRA Model.
6-9           The parametric uncertainty analysis as                   FQ-A4         Step not performed.                              This F&O has been resolved.,
discussed in QU-E3 (estimate of uncertainty              QU-A3         Perform the FPRA uncertainty analysis,           Parametric uncertainty has been performed intervals, etc.) is not performed. Also, the                           including estimates of uncertainty bounds,       for CDF and LERF for each unit's FPRA.
               'state-of- knowledge" correlation between                              per the requirements of QU-A and QU-E.
fire-specific event probabilities (e.g.,
suppression       system unavailabilities, ignition frequencies,                         fire hot short conditional                      When   performing   parametric  uncertainty calculations, ensure   uncertainty intervals for ignaitiontfrequnies,e     hasnot short cononapld                       event probabilities utilized by the FPRA are probabilities,     etc.) hasn't (This F&O originated         fromyet SRbeen  applied, QU-A3)                          correlated when significant.
7-1           A review of the quantification results for               PRM-A3        The method in which the fire-induced             This F&O has been resolved.
selected compartments involving fire-                    PRM-A4        spurious safety injection actuation and          A review of the model was performed and induced safety injection actuation and fire-             PRM-B5        spurious opening of atmospheric dump             revision made to address and resolve the Revision 0                                                                                                                                                                Page V-61

Florida Power & Light                                                                                                                             Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                       Discussion                       Fire PRA SR             Basis and Recommendation                           Disposition in Fire PRA Update induced opening of atmospheric dump                 PRM-B9        valve initiating events were linked into the           issue identified in the F&O. Additional valves was performed to verify that the                           fault tree produces conservative results that          reviews were performed as part of the modeling was consistent with the internal                         could impact the determination of significant          overall results and cutset reviews and no events PRA treatment of similar initiating                        contributors to fire-induced risk.                     additional instances were identified.
events. This review revealed that duplicate                        Review the quantification results for the fire-cutsets were being introduced by the                              induced initiating events to verify that the manner in which the new logic for capturing                       results are consistent with the comparable the fire-induced initiating events was linked                     internal events.
into the fault tree. For example, a review of                     Review the mapping of the fire-induced cutsets for zone 098-A showed that the top                        initiating event impacts to ensure that they two cutsets were identical except that one                        i ni          stent i th t ompare th a l used bleed aand version feed of  the HFE based      for alignment on reactor  trip     of                 are eventsconsistent initiator, wit thate that    comparable appropriate       internal differences bleedrinndifeedGbasedevnlrendtor          orip rdue                     to the considerations of the fire PRA are occurring with SG low level and the other                         incorporated (e.g., application of bounding used a version of the HFE based on timing                         timing for HEPs to capture uncertainty in the associated with trip with nominal SG level.seuneofi-ndcdalrs)ortt Similar for zoneissues   were identified in the cutsets 091-ETL.                                               sequence deviations of  in fire-induced the modelingfailures),    or that are documented and justified.
As noted in the 2010 peer review in F&O 1-                         Review the application of the feed and bleed 4, there are also inconsistencies in modeling                     HFE to ensure the appropriate timing is used of the fire-induced small LOCA when                               during the fire quantification. Since the compared to the internal events small LOCA                        MFW pumps are assumed failed for all fire initiating event. logic due to circular     PTNissues, explains andthat  this was a sensitivity                MFW areas, pumoste              a the most appropriate   te vale   may be value may     be case sos tiscrtogcbues      n  a  d signstifnty                   the   HEPwith based case shows this to be a non-significant                           occurs           low on  timing  assuming level in the SGs.       the trip issue. However, it is not clear that the circular logic issue could not be resolved                         Review treatment of any additional HEPs and that all potential impacts of the modeling                    with event-specific timing assumptions to approach taken are understood.                                    ensure that the appropriate values are used (This F&O originated from SR PRM-A3)                               in the fire quantification.
7-3            The current model uses the LERF model for           PRM-B14        It cannot be determined from the existing             This F&O has been resolved.
the PTN revision 9 model (PTN-BJFR                            documentation that an assessment was                   A review of the mapping of Level 1 010, Rev. 1) and maps appropriate                                  performed to identify new accident                    sequences to the plant damage states in the equipment impacts into the system models                           progressions beyond the onset of core                 LERF model was reviewed. No new used to model LERF. No new accident                               damage that would be applicable to the Fire            accident progressions that required progressions beyond the onset of core                              PRA that were not addressed for LERF                  modification of the LERF model were damage were identified for the fire PRA.                          estimation in the Internal Events PRA.                 identified.
However, there is no documentation that a                          Document an assessment to determine if specific review of the accident progressions Revision 0                                                                                                                                                                   Page V-62

0                                                                           0 Florida Power & Light                                                                                                                     Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR              Basis and Recommendation                       Disposition in Fire PRA Update leading to LERF was conducted to identify                      there are potential fire-induced LERF whether new considerations should be                            mechanisms not captured by the internal addressed in the fire PRA.                                     events accident progression models. If none In addition, effects on PDS mapping due to                     are identified, document the basis of that fire-induced failures may not be                               conclusion.
appropriately captured. For example, RWST diversion of the RWST to the containment sump is modeled as a failure of HHSI which would normally go to a dry containment PDS. However, the actual PDS should be one for wet containment. While this is a late containment failure concern rather than a concern for LERF, there may be similar fire-induced failures that could affect the mapping of LERF accident progressions.
(This F&O originated from SR PRM-B14) 7-6           The new fire-specific safe shutdown actions     HR-El           Final post-fire safe shutdown actions have        This F&O has not been resolved.
which are credited in the final Fire PRA will   HR-E2           not been defined and appropriate                  The FPRA includes various actions that are be proposed to be added to the plant fire                       procedures revised to include the actions to      being included as required plant changes in response procedures. These human actions        HR-E3           be credited in the Fire PRA.                      the NFPA 805 LAR. The development and are included in the ALTEREDEVENTS table         HR-E4           Complete the identification of new fire-          implementation of related procedures has of the FRANC model using component basic        HR-H2           specific safe shutdown actions which are          not yet been initiated as it is part of the events as surrogate.                            HR-I1           credited in the final Fire PRA and evaluate       overall integrated process associated with However, the safe shutdown actions              HR-12           and document the HEPs consistent with             transition to an NFPA 805 license basis.
modeled in the FPRA are not currently                           processes used for internal events HEPs.
consistent with those specified in the plant    HR-13           Include consideration of fire effects on the fire response procedures, there is no           HRA-A2          operator action, availability of cues, documented assessment of the cues               HRA-A4          availability of time to complete the action, required to initiate the actions, no training                   feasibility of the credited actions given a fire, has been provided to operators on the new       HRA-B2          and potential 92-18 impacts for both fire-specific actions, no operator reviews or   HRA-B3          screening values and detailed HEP talk-throughs of the credited actions has       HRA-D2          development.
been documented, and the applicable             HRA-E1          Also, complete operator reviews and/or talk-performance shaping factors have not been                       throughs when the procedure updates are considered, including time available for the    PRM-B311        completed to ensure that the interpretation action.                                                         of the actions is consistent with the This F&O supersedes 2010 Peer Review                            operator's understanding and training.
F&Os 1-41, 2-6, 6-4 and 6-11.                                   Finally, consider expanding the discussion Revision 0                                                                                                                                                             Page V-63

Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR            Basis and Recommendation                  Disposition in Fire PRA Update (This F&O originated from SR HRA-A2)                           of sources of model uncertainty related to the HRA to include consideration of the accuracy and completeness issues noted in NUREG-6850, Volume 2, Appendix V.
7-8            Dependency between multiple altered             HR-H3          The dependency associated with operator      This F&O has been resolved.
events representing new HFEs in the same        HR-12          actions applied using the altered events     The use of altered events as a surrogate for cutset and between action represented by        HRA-D2         method has not been addressed.               a recovery action has been significantly the altered events and other HFEs in the                       Address dependency between multiple          reduced as noted previously. Those same cutset has not been assessed based         PRM-B11        altered events representing new HFEs and     remaining instances are addressed by on the assumption that the dependency                          between the altered events and other HEPs    modifications to the recovery rule file so that effects are bounded by the application of                      in the same cutset. If detailed dependency   only a single instance of this use would exist conservative screening values. However,                        analysis is not performed, provide a         in any cutset. This eliminates the potential there is no documented assessment to                           justification supporting the assumption that for multiple surrogate recovery events to support this assumption. There are cases                       the values chosen for the altered events     appear together in the same cutset.
where complete dependency between                              bounds dependency effects.
events may be appropriate. For example, cutsets 40 - 45 in the provided Aggregate CDF-aggregate.cut file contain altered events MAVC4200A_1.OOE-01 and MAVC4460_1.OOE-02 in each cutset. The product of these two events is therefore 1.OOE-03. However, since both events involve failure to isolate the letdown line, it could be assumed that there is complete dependence between the events since they would share a common cue.
This F&O supersedes 2010 Peer Review F&O 6-16.
(This F&O originated from SR HR-H3) 8-3            Attachment U - Internal Events PRA Quality      PRM-B2         The potential effect of internal events F&O  This F&O has been resolved.
(DRAFT), document applicability of Internal                    disposition on development of the FIRE PRA   The internal events PRA model F&Os that Events F&Os to internal events PRA, but not                    was not addressed.                           have not been resolved/closed have been to Fire PRA. There was no evidence that                        Review internal events F&Os and provide      reviewed and found to have no negative the review of F&O disposition status                           documentation as to how disposition of       impact on Fire PRA results or this addressed the question of whether the                          those F&Os may impact development of the     application.
disposition that was taken would adversely                     Fire PRA.
affect the development of the fire PRA.
Revision 0                                                                                                                                                      Page V-64

0 Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR            Basis and Recommendation                     Disposition in Fire PRA Update This F&O is derived from 2010 Fire PRA peer review F&O 4-4.
(This F&O originated from SR PRM-B2) 8-5            The Fire PRA model changes were                HRA-B1         The standard requires a review of the fire-    This F&O has been resolved.
constructed so as to allow credit for the      HRA-B3         induced initiating events accident current internal events PRA model structure                   sequences, and success criteria included in    The review of results and cutsets that were using existing accident sequence                              the internal events model, to identify new progression, success criteria and timing.      PRM-B7         accident sequence progressions or success      performed did identify a number of instances The internal events HRAs are modified with                     criteria due to unique aspects of fires,       such as that specifically identified in the F&O. In all instances, it was determined that a screening modifier. Travel paths are                        This review will help assure that there are no the existing model structure was appropriate considered in the human failure evaluation                    revised actions where the screening            and that opportunities for recovery actions report.                                                       multipliers are not appropriate,               were limited either because of a lack of However, there is no indication that a review                 Conduct and document a review of the fire-     appropriate cues or insufficient timing to gain was performed to identify accident                             induced initiating events accident             any meaningful benefit via recovery. The sequences that may require modification                        sequences, and success criteria included in    analysis documentation of the HFE based on unique aspects of the plant fire                     the internal events model, to identify new     treatment was updated to address the response procedures. For example, RWST                        accident sequence progressions or success      internal events PRA model human actions draindown may affect the evaluation of                         criteria due to unique aspects of fires,       that are used in the FPRA. The timing for aligning sump recirculation, which                                                                 documentation addresses the applicability, is not presently represented in the non-                                                                     numerical adjustment, and availability of LOCA transient event tree accident                                                                           necessary cues.
sequences used for the majority of the fire scenarios.
A review should be performed for possible changes to success criteria, particularly due to model changes from the MSO evaluation.
This F&O is derived from 2010 Fire peer review F&O 6.3.
(This F&O originated from SR PRM-B5) 8-8           Several portions of the analysis are not        FSS-E3         Unit 3 results not fully documented, although  This F&O has been resolved.
documented for Unit 3. Specific examples        PRM-A3         they are available for inspection using the    The U3 results have been added to the include:                                        PRM-C1         quantification software.                       analysis documentation.
Unit 4 significant contributors are identified                 Document Unit 3 results consistent with the in 0493060006.005, Rev. 2. Unit 3                              Unit 4 results.
significant contributors are available, but not fully documented.
Revision 0                                                                                                                                                        Page V-65

Florida Power & Light                                                                                                                 Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR            Basis and Recommendation                     Disposition in Fire PRA Update Unit 4 Fire Scenario information is presented in 0493060006.004, Rev. 2, but the equivalent Unit 3 information (Attachment D and E) is not provided.
(This F&O originated from SR PRM-A3) 8-10           2010 FPRA peer review F&O 1-44 finds           HRA-Cl         This appears to have a significant non-        This F&O has been resolved.
issues with masking fire effects by setting    PRM-B1 1       conservative impact to PRA results. Given      The use of '0'; has been eliminated in the basic events to 0 in the Altered Events table.                the actions in the altered events report are   Altered Events table. Instead, events are This has partially been addressed by setting                  being added to the model as needed             set to nominal. In the case of the application these events to 'nominal' in the Altered                      recoveries in order to ensure risk is low, and calculations for NFPA 805, the 'compliant' Events table and for reviewing cases where                    given the resulting recovery actions do not    case is determined by using a '0' value the nominal value is on one side of an AND                    show up in the results in most cases, there    which would under-estimate the compliance gate and the modified HEP value is on the                     appears to be a disconnect between the         case risk and thereby provide a conservative other side. However, there are still cases                    addition of new actions to the procedures      estimate of the risk increase for the where fire impacts are masked when the                        and the quantification of these actions in the application.
nominally adjusted event is on both sides of                  FPRA. It appears part of the disconnect is an AND gate or the HEP event is on one                        that the logic modeling, as modified by the side of an AND gate and nominally adjusted                    altered events table, results in the recovery events are on the other side. Scenarios 030                   values being screened from the results.
PTB and 067E PTB are two examples. This                       Due to the complexity of this methodology, it old F&O is converted to a new F&O 8-10.                       seems a difficult task to review and address (This F&O originated from SR HRA-Cl)                          for these masking issues. Perhaps a more systematic and comprehensive approach, with an independent review, could provide confidence that these nonconservatisms are addressed. Adding new HEP basic events, consistent with the approach used for internal events, would address this issue.
9-1            A general screening based on the ability to    FSS-G2         No basis for the screening criteria is         This F&O has been resolved.
form a damaging HGL in an exposing             FSS-G3         described. Since no MCA scenarios are          The existing HLG/MCA analysis includes a compartment was developed. If no                              developed, there is no way to determine if     number of occurrences where the simplified damaging HGL could form in an exposing         FSSG6          the exceeded (yet applied) screening criteria  screening approach was found to generate compartment then there was no possible                        are significant.                               over-conservative results. Incrementally associated MC scenario.                                       Provide a basis for the 1 E-07/yr screening    enhanced treatments were applied to When a damaging HGL could form, a                             criteria including additional information      confirm that these locations had a very low second screening was performed whereby                        required when the criteria are included. The   likelihood of creating or causing formation of the frequency of developing the HGL was                       basis for the screening criteria should        HGL conditions and consequently a possible Revision 0                                                                                                                                                        Page V-66

0 Florida Power & Light                                                                                                                      Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                     Fire PRA SR             Basis and Recommendation                       Disposition in Fire PRA Update determined. If the frequency was less than                       ensure the frequency is not too high, thereby    multi-compartment scenario 1E-07/yr, then the scenario could be                             potentially masking significant MCA eliminated. However, there was no basis                          scenarios.
provided for the 1E-07/yr criteria, nor was                      Evaluate impact of exceeded screening the criteria adhered to; in fact, the criteria                   criteria; for example, a qualitative analysis of was exceeded, yet still applied, in over 150                     the expected CCDP based on known targets different scenarios. Some of the screened                        in the exposing and exposed compartments.
scenarios were slightly over the 1E-07/yr threshold, while others ranged as high as nearly 6E-07/yr.
The impact of exceeding the criteria cannot be determined as no specific MCA scenarios were ever developed; therefore, it is not known if the scenarios would be significant.
For example, if it is assumed that 10 of the scenarios with a frequency of 5E-07 had CCDPs of 1.0, this would result in an increase in total CDF of 5E-06 which is about 10% of the total fire CDF.
(This F&O originated from SR FSS-G2) 9-4            The multi-compartment analysis assumes a         FSS-G4          Systematic generic assessment of active fire     This F&O has been resolved.
bounding value of 7.4E-3 for evaluation of       FSS-G5          barrier elements may lead to non-                The update of the analysis to incorporate a active fire barrier elements. Actual fire                        conservative results.                            barrier failure probability that integrates all barrier elements are not considered; instead                     Ifa screening value is desired, NUREG/CR-        possible barrier elements was found to result the failure probability of a fire door is                        6850 Section 11.5.4.4 suggests using a           in a value of approximately double the assumed for active barrier element failure                       screening value of 0.1 for active fire barrier   current value. However, since the entire because this failure probability represents                      elements. This value is much more likely to      analysis approach involves a screening the highest single probability of a single                        encompass multiple fire barrier elements.        strategy, additional analysis refinements are barrier failure. This method ignores the                         For scenarios that do not screen out, actual     possible. An assessment of the use of a potential for multiple fire barrier elements.                    fire barrier elements identified during          higher barrier failure probability to account Per NUREG/CR-4840 (source document for                           walkdowns (or document review) can be            for failure of multiple barrier elements found NUREG/CR-6850 Table 11-3, "Barrier                               used to develop a more realistic barrier         that the overall conclusion that MCA Types and Their Failure Probabilities".) the                      failure probability,                             scenarios are not risk significant and need total barrier failure rate is a union of the                                                                       not be explicitly included in the FPRA was probabilities of the individual failure rates.                                                                    confirmed. However, the analysis Therefore, a value of 7.4E-03 may be                                                                               documentation has not yet been updated ot conservative or non-conservative.                                                                                 reflect these results and insights.
This is based on 2010 FPRA peer review Revision 0                                                                                                                                                              Page V-67

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR             Basis and Recommendation                     Disposition in Fire PRA Update F&O 1-35.
(This F&O originated from SR FSS-G5) 9-5            The screening criteria are defined in the      FSS-G2          Use of a standard fire scenario may be non-    This F&O has been resolved.
Turkey Point Hot Gas Layer and Multi-                          conservative for some zones.                   The use of the 5 minute delay to Compartment Analysis, (Report                                  Review zones to ensure that the standard       combustible cable ignition is considered H0493060006.006) methodology,                                  fire scenario is actually the most challenging realistic. Other conservatisms in the Compartments that don't screen are retained                    scenario inherent to the analyzed              analysis ensure the overall conservatism of for further analysis.                                          compartment. For zones where the               the MCA/HGL evaluation.
A concern identified with the screening                        standard fire scenario is not the most criteria involves the use of a standard fire                   challenging, determine the most challenging scenario for each analysis rather than                         scenario and evaluate accordingly.
determining the most challenging fire scenario inherent to the analyzed compartment. This approach potentially masks the potential for forming an HGL in the exposing compartment.
For example, in zones 67 and 68 the standard fire scenario is non-conservative due to the potential for HEAF in 4kV switchgear. The damage time of 5 minutes is non-conservative for HEAF scenarios (should use 0 minutes).
This F&O supersedes 2010 FPRA peer review F&O 3-11.
(This F&O originated from SR FSS-G2) 9-6            The system unavailability records for the      FSS-D7          This is a systematic issue. The intent for     This F&O has been resolved. The fire plant have not been reviewed in crediting                      Capability Category II is to additionally      protection system availability data for PTN fire detection and suppression systems.                        require a review of plant records to           has been reviewed and no outlier behavior This F&O supersedes 2010 FPRA peer                             determine if the generic unavailability credit has been identified.
review F&O 2-26                                                is consistent with actual system unavailability. Outlier experience would be (This F&O originated from SR FSS-D7)                           any experience indicating that actual system is unavailable more frequently than would be indicated by the generic values.
Consider performing and documenting the review of plant records to determine if the Revision 0                                                                                                                                                        Page V-68

Florida Power & Light                                                                                                                    Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR              Basis and Recommendation                      Disposition in Fire PRA Update generic unavailability credit is consistent with actual system unavailability. Outlier experience would be any experience indicating that actual system is unavailable more frequently than would be indicated by the generic values.
9-10           Section 3.1 of the FSS Report                  FSS-A6          Assumption made that no cabinet/panel fires      This F&O has been resolved.
(0493060006.004, Rev. 2) states: "For the      FSS-H7          in the MCR will ever spread to an adjacent       Panels with communication between electrical panel fires, the scenarios are                       cabinet even if the cabinets are open to one     adjacent panels are to be provided with developed similar to scenarios involving                        another. This incorporates an implied            incipient detection to ensure early electrical panel fires outside the Control                      assumption that every MCR panel/cabinet          identification of fire to preclude spread Room and are adequately described in                           fire will be extinguished prior to spread.       between panels.
Attachment A. Fire spread to adjacent                           Identify adjacent MCR cabinets/panels panels was determined for these scenarios                       which could result in fire spread given failure based on a walkdown of the control room                         of suppression. Apply NUREG/CR-6850 during which panels with potential barriers                     Appendix L, S or other relevant document to for spread of fire were opened to confirm the                   address the potential for fire spread.
existence of such barriers. For MCB fires,                      Ensure documentation is consistent with the the method from NUREG/CR-6850                                    proc         ually us in           the Appendix L is applied. NUREG/CR-6850                            process actually used in the analysis.
Appendix L defines a non-suppression probability applicable to the MCB. From Figure L-1 of NUREG/CR-6850, for non-qualified cables, and for a bounding distance of 0 meters (assuming that the cables terminating at the individual MCB are in very close proximity), a non-suppression frequency of 8.30E-3 is used for the MCB."
However, based on discussion with FPL/ERIN staff, this was not done.
Essentially, no fire spread for any cabinet in the MCR was assumed. For panels with incipient detection, success of the detection results in no damage as it is assumed operators isolate the circuit prior to additional damage in the cabinet. If insipient detection fails, the MCB panel fails completely, but never spreads another cabinet.
Revision 0                                                                                                                                                            Page V-69

Florida Power& Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                    Discussion                      Fire PRA SR            Basis and Recommendation                       Disposition in Fire PRA Update For all other cabinets/MCB panels without incipient detection, full burnout of the cabinet/panel is assumed, but again, no spread to adjacent cabinets is assumed even if the cabinets are open to one another (e.g., walkthrough MCB).
(This F&O originated from SR FSS-A6) 9-11          Several entries in Table 3-1 of Report           FSS-G2         It was not confirmed during the peer review      This F&O has been resolved.
H0493060006.006 says 'Valkdown required                         that this was just a documentation issue.        It was confirmed that the lower damage to confirm no combustibles within the 383                       Therefore, this is classified as a finding       threshold associated with thermoplastic Zor"when the cables are not IEEE-383                            because it could affect the analysis results. materials was used for the analysis. The qualified. Based on discussions with FPL                        Verify that the damage criteria used is          typographical error has been corrected.
and contractors, it is believed that this is a                  consistent with non-383 cable damage and typo, and the correct damage criteria were                      revise the documentation as required. If it is actually applied.                                               discovered that the incorrect damage criteria (This F&O originated from SR FSS-G2)                            were applied, update the analysis with the correct criteria.
10-1          The 2010 peer review identified that "Fire       FSS-Cl         The present analysis provides a bounding         This F&O has been resolved.
modeling was conducted via generic fire          FSS-G1         approach for fire severity in most cases,        The recommended resolution action in the modeling from which Zones-Of-Influence                          since the 98th percentile fire heat release      F&O was assessed in the context of the (ZOI) for specific initiator types was                          rate is used.                                    dominant fire risk contributors. This generated. The ZOls were used to define                         However, use of the split fraction method is     assessment concluded that further bounding fire characteristics for each fire                     based on industry events rather than site        refinements such as that described in the scenario. Characteristics that are used to                      specific fire ignition sources and target        F&O would not substantively change the bound potentially risk contributing fire events                 configurations. Therefore, this could result     results of the analysis. The existing are identified in Attachment B of the Fire                      in non-conservative frequency estimates of       treatment retains some conservatism which Scenario Report, (Report 0493060006.004).                       target damage.                                   results in this SR meeting CC I. This is Based on the use of a bounding approach                         Perform 2-point fire modeling, when              adequate for the NFPA 805 application, as this SR is judged to be met at CC 1.                            applicable, for risk significant fire scenarios, this conservative bias would tend to over-Significant fire scenarios should be                                                                             estimate the risk metric that is used to judge developed with 2-point fire modeling."                                                                           the acceptability of this application.
Since this review, FP&L has stated that "The                                                                     The issue regarding the ERIN panel split use of a panel split fraction to differentiate                                                                   fraction is addressed in the disposition for between fires impacting the panel and                                                                            F&O 10-3.
components with cables terminating at the panel versus panel fires impacting cables outside of the panel provides an equivalent Revision 0                                                                                                                                                             Page V-70 0

0 Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR            Basis and Recommendation                    Disposition in Fire PRA Update and more useful two point fire model."
The Panel Split fraction is developed from a supplemental report (ERIN report, Supplemental Fire PRA Methods, dated February 2010). This document was submitted to the EPRI Fire PRA Methods Review Panel. This review is not complete as of the date of this peer review.
Use of the split fraction method is based on industry events rather than site specific fire ignition sources and target configurations and therefore, could result in non-conservative frequency estimates of target damage.
(This F&O originated from SR FSS-C1) 10-2          The 2010 review of PTN Tasks 8 and 11           FSS-A1          Including the fire frequency and associated  This F&O has been resolved.
Report 0493060006.004, identified that 'no                     fire scenarios from hydrogen fires will have Miscellaneous hydrogen fires have been hydrogen fires other than turbine/generator                    impact to the CDF and LERF results.          incorporated in the Fire PRA in the charging have been postulated.' (Previously F&O 5-                      Incorporate the hydrogen fire scenarios      pump room fire areas where the hydrogen
: 16)                                                            being developed into the model, and update   lines associated with VCT cover gas are Since this Finding was identified, FP&L has                     documentation as necessary.                  routed.
determined that 'Miscellaneous Hydrogen piping at PTN is limited to hydrogen supply to the VCT tanks. The associated piping is located in the charging pump rooms (Fire Zones 45 and 55). Fires in these fire zones are assumed to impact all components in the fire zone. The associated risk is low given the availability of thermal barrier cooling for RCP seals and HHSI pumps. Allocation of the IGF associated with miscellaneous hydrogen fires to these fire zones would result in an increase in the ignition frequency for these zones by less than a factor of 3.
Given the low risk significance of these zones this will have a negligible impact on overall plant risk and the charging pump rooms will remain low risk contribution fire Page V-71 Revision 0                                                                                                                                                     Page V-71

Florida Power & Light                                                                                                                 Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                   Fire PRA SR            Basis and Recommendation                      Disposition in Fire PRA Update zones.
Incorporation of this ignition frequency into the associated documentation will be incorporated in a future revision to the documentation.'
Hydrogen fires are also being developed for H2 piping and valves in Compartments 82 and 87 (scenarios 82-P and 87-P).
However, since these do not appear yet in the Fire Scenario Report, action is required.
This finding is currently being addressed and appears to be resolved once the new H2 fires are included in the model and documentation is updated.
(This F&O originated from SR FSS-Al) 10-3           FSS-C4 requires severity factors to be         FSS-C4          Severity factor (panel split fraction) is used The FPRA quantification uses the panel independent of other factors. Fire severity    FSS-D5          extensively in the Fire PRA.                   factors consistent with the latest guidance factor as discussed in Section 7.1.2 for                       Use the severity factor method described in    from the EPRI Methods Review panel. A electrical cabinets is not developed or        FSS-G1                                                         sensitivity study has been performed to NUREG/CR-6850, or develop an accepted applied consistently with the NUREG/CR-                        industry approach (presently being             address the impact of elimination of the 6850 methods. This is developed from a                         discussed by EPRI). Develop fire severity      credit for the panel factors. The results of supplemental report (ERIN report,                              factors based on the likely HRR and location   this evaluation indicate that the delta Supplemental Fire PRA Methods, dated                           of overhead cables or location of equipment. CDF/LERF would exceed the Reg Guide February 2010). This document was                              For example, if cable is 7 feet overhead, the  1.174 guidelines should these factors be submitted to the EPRI Fire PRA Methods                         severity factor would be based on the          completely eliminated (the 1E-5/1 E-6 delta Review Panel. This review is not complete                      minimum HRR that would damage the cable        CDF/delta LERF limits would be exceeded as of the date of this peer review.                            at that distance. Additionally, the growth     but the conservatively calculated delta risk Using this method, fire propagation outside                    time can be used in determining non-           would be less than 2E-5/2E-6). Further of the electrical cabinets is dependent on the                 suppression time.                              refinements of this sensitivity evaluation are nonsuppression probability. Therefore,                                                                        possible to reduce the calculated delta risk.
some dependency exists in this data if used                                                                   Credit for additional Defense In Depth in conjunction with a non-suppression factor.                                                                 measures may be taken in areas of concern Due to this derivation of the conditional                                                                     as necessary to compensate for the probabilities for fire propagation outside of                                                                 increased delta risk.
the cabinets, the conditional probabilities thus developed (and applied in the FRANC model) could potentially be non-Page V-7'2 Revision 0 Revision                                                                                                                                                           Page V-72

Florida Power& Light                                                                                                                 Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                   Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update conservative.
The severity factors are developed using generic fire events data from the EPRI fire events database. Given the fire data duration and damage is a result of multiple factors (growth, suppression, severity, location, etc), and given the fire data often does not have sufficient information to make a reasonable determination of either the fire size or whether a fire propagated outside the cabinet, the severity factor used (panel split fraction) may not necessarily bound the conditions of the specific fire scenarios under analysis.
(This F&O originated from SR FSS-C4) 10-4          One situation was identified for which credit  FSS-C8          This finding is based on identification of    This F&O has been resolved.
of fire wrap is taken in Compartment 96 for                    credit for a wrap in Attachment A of the Fire A qualitative assessment has been ignition source 3B04, which is a 480V load                     Scenario Report, (Report 0493060006.004). performed to assess the potential impact of center. This fire wrap protects PB3319,                        Any credited fire wrap should be addressed    this F&O.
PB3813, PB7022, and P17521. The wrap                           and the wrap integrity should be established appears as being credited in a HEAF                                                                          The hose stream test imposed on the fire with respect to fire resistance, mechanical   barrier qualification subsequent to fire scenario. No justification for crediting this                  protection, and potential fire related wrap assuming mechanical damage and                                                                          exposure is considered to provide a exposure to which the wrap may be exposed     comparable level challenge to the thermolag direct flame impingement from the HEAF is                      (direct flame impingement, HEAF, etc.).
provided. Similar issue for 3B03 also in                                                                     barrier as would the HEAF force applied at Compartment 96.                                                                                              the onset of fire exposure.
Thermo-lag is also seen as credited in some scenarios, which would require justification due to issues with this particular type of cable barrier.
(This F&O originated from SR FSS-C8) 10-6          Treatment for transient fire damage to         FSS-D6          Many transient fire scenarios have been       This F&O has been resolved.
targets is measured from the compartment                       screened during detailed scenario analysis. Supplemental walkdowns were performed to floor rather than the height of the transient                  The results of the FPRA are therefore         re-assess the treatment of transient fires.
fuel package that is typically considered.                     potentially non-conservative for the analyzed These walkdowns focused on two key Discussion with FP&L during the review                         detailed scenarios.                           attributes - the appropriateness of the provided some basis for the damage height                      The transient fires should be considered to   selected HRR characterization and the Revision 0                                                                                                                                                      Page V-73

Florida Power & Light                                                                                                             Aftachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                       Discussion                    Fire PRA SR         Basis and Recommendation                 Disposition in Fire PRA Update (indicating that transient fires above the floor               be above the floor level in the analysis. location of the postulated fire scenarios.
will have an overall lower average surface                                                               With respect to this specific F&O, the HRR). However, the supplemental                                                                          placement (elevation) of the assumed fire discussion was still considered inconsistent                                                             was based on the physical features of the with past events and existing guidance on                                                                location. The fires were not artificially analysis of transient fires, and could lead to                                                           elevated in the absence of a physical non-conservative estimates of transient fire                                                             feature.
damage to targets.
Transient fire evaluations conducted as described in the Fire Scenario Report result in screening fire damage to targets that are located > 7.3' above the floor which is believed to be non-conservative for developed fires involving ordinary combustible fuel packages such as a trash can or trash bag. In response to this concern it was pointed out that the thermal plume component relies on empirical relationships between the source strength and the distance between the virtual origin of the fire and the target. The fire plume begins to entrain air at the lowest point of burning, which defines the base of the fire; normally at the floor. However this argument ignores the potential that a fire could begin burning at the top of a fuel package thus elevating its base. At a minimum, during the initial period of burning, damage temperatures generated by the fire would likewise be elevated. Over time the base of the fire may change due to collapse of the fuel package or burning away of the fuel, however the empirical model presented did not present sufficient basis for assuming that the base of the fire is at the floor for its entire duration.
(This F&O originated from SR FSS-D6) 10-8           Ambient conditions are assumed in the            FSS-D4        Underestimating the ambient conditions    This F&O has been resolved.
Revision 0                                                                                                                                                    Page V-74

0 Florida Power & Light                                                                                                                     Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                       Fire PRA SR            Basis and Recommendation                      Disposition in Fire PRA Update Generic Fire Modeling Treatment Report            FSS-H4         could result in non-conservative estimations   A qualitative assessment has been (prepared by Hughes). Ambient                                    of zones of influence and targets considered   performed to assess the potential impact of temperature is assumed to be 68&deg;F for all                        to be fire damaged.                            this F&O.
calculations. No technical discussion or                         Assess areas where elevated ambient            The sensitivity of the ZOI dimensions to the justification is provided in the Fire Scenario                    temperatures could be experienced and          ambient temperature is relatively low as Report to substantiate that this is a                            justify the acceptability of the models used. described in the original Hughes Generic reasonable value for the compartments                            Otherwise, incorporate elevated ambient        Fire Modeling treatments report, in particular where this was applied.                                          temperatures into the zone of influence        for IEEE-383 qualified/Thermoset cables. In (This F&O originated from SR FSS-D4)                             calculations,                                  the case of an initial ambient temperature of 35&deg;C, the expected affect on the ZOI dimensions is within the measurement uncertainty in the field.
10-9           The 2010 peer review identified that 'Except      FSS-D8         The method currently employed could result     A qualitative assessment has been for the MCR fire scenarios, no other fire                        in optimistic times for suppression activation performed to assess the potential impact of scenario has used the Non-Suppression                            or fire brigade response.                      this F&O.
Probability (NSP) in PTN fire model at this                      Assess and document the effectiveness of       The HGL and MCA analyses credit both time.' Since this review, FP&L has taken                          suppression with respect to:                   automatic suppression system and fire manual) in the Multi-Compartment/Hot Gas                         System design complies with applicable         brigade actions. In the context of the HGL mayerual)intheMuatio.oweverthsevaution     Gcodes                        and standards, and current fire        and MCA, the fire brigade action of interest Layer evaluation. However, this evaluation                       protection engineering practice,               is fire control as that would terminate the does not include an assessment of the fireposbltfr                                                                                 Gfrmin.Hwvth protection system effectiveness.                                 The time available to suppress the fire prior  only readily available numeric credit is fire Of particular concern is that fire detection                      to target damage,                              suppression credit. To reduce the and/or suppression timing (i.e., thermal                          Specific features of physical analysis unit    conservatism introduced into the analysis, response of the detector and/or sprinkler)                       and fire scenario under analysis (e.g.,        fire detection time is ignored for the HGL was not calculated and subtracted from the                        pocketing effects, blockages that might        and MCA. The timeframe associated with time considered for manual suppression                            impact plume behaviors or the "visibility" of  detection and suppression is significantly when using the FAQ-0050 process. In                               the fire to detection and suppression          less than the timeframe required to reach a addition, fire detection reliabilities are not                    systems, and suppression system                hot gas layer temperature which would included in the assessment. If the detection                      coverage), and                                 impact the HGL analysis.
system does not function as intended, the                         Suitability of the installed system given the time to detection to initiate fire brigade                        nature of the fire source being analyzed.
response would be substantially longer.
(This F&O originated from SR FSS-D8) 10-11         The 2010 peer review identified that "fire         FSS-C2         The present analysis provides a bounding       This F&O has been resolved.
scenario evaluation tools were developed           FSS-C3         approach in most cases, since the 98th         The recommended resolution involves the based on the Generic Fire Modeling                                percentile fire heat release rate is used from crediting of growth and decay in the Revision 0                                                                                                                                                           Page V-75

Florida Power & Light                                                                                                                  Affachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR              Basis and Recommendation                    Disposition in Fire PRA Update Treatments. These walkdown/evaluation           FSS-G1          fire initiation without growth and burnout,    modeling of the postulated fire. The existing tools are based on bounding fires that are                      However, use of the split fraction method is   analysis does not take credit for these assumed to cause target damage at a height                      based on industry events rather than site      variables. A review of the dominant fire above the base fire with the fire burning at                    specific fire ignition sources and target      scenarios found that the risk benefit that peak intensity and without burnout times.                       configurations. Therefore, this could result   might be gained is minimal. Therefore, this Because these tools assume a fire burning                       in non-conservative frequency estimates of     refinement was not performed. The at peak intensity and without burnout, this                     target damage.                                 resulting categorization of the related SR is SR is consideredSRiscosierdme a C met at CCI"I"Include                        Irisk fire growth and decay for rikconservatismCC 1. Since thebeing  retainedresults approach      in the in results, some Since the review, FP&L has stated that "The                     significant fire scenarios.                    thiseCCaisu       be     aeqate       the fo use of a panel split fraction to differentiate                                                                 this CC is judged to be adequate for the between fires impacting the panel and                                                                          NFPA 805 applications as the conservative components with cables terminating at the                                                                      bias would tend to result in the over-panel versus panel fires impacting cables                                                                      estimation of the risk metrics used for this outside of the panel provides an equivalent                                                                    application.
and more useful two point fire model... The application of the two point treatment to individual fire scenarios is carried through to the MCA/HGL evaluation which addresses the impact of each scenario on MCA."
The Panel Split fraction is developed from a supplemental report (ERIN report, Supplemental Fire PRA Methods, dated February 2010). This document was submitted to the EPRI Fire PRA Methods Review Panel. This review is not complete as of the date of this peer review.
Use of the split fraction method is based on industry events rather than site specific fire ignition sources and target configurations and therefore, could result in non-conservative frequency estimates of target damage.
(This F&O originated from SR FSS-C2) 10-12          The 2010 peer review identified that "The       FSS-D9          This appears to be a documentation issue,      This F&O has been resolved.
PTN FPRA methodology generally does not                         but FP&L should confirm that smoke             An analysis of the impact of smoke damage include postulation or evaluation of smoke                      damage has been considered and document        has been completed and documented in the damage. Additional review shows that the                        accordingly.                                   PTN FPRA Scenario Report.
smoke issues do not affect the FPRA results                     Confirm that smoke damage has been Revision 0                                                                                                                                                            Page V-76

Florida Power & Light                                                                                                               Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                      Discussion                    Fire PRA SR            Basis and Recommendation                  Disposition in Fire PRA Update significantly. However, the FPRA does not                       considered and document accordingly.
include a qualitative evaluation of smoke damage to FPRA equipment."
Since the 2010 review, FP&L stated, "Section 6.2 of the Scenario Report was added to address this concern." However, section 6.2 provides a high level discussion and methodology including the statement that "Exposure time plays a key role in the likelihood of failures from smoke. As a result, damage from short term smoke exposure will only result from severe conditions.... Instruments, control components and all high voltage powered components are exceptionally vulnerable to circuit bridging as a result of airborne smoke and deposited particulates."
However, there is no documented discussion of the smoke damage assessment results, and none of the targets in the scenarios indicated smoke damage as the failure mode. NUREG/CR-6850 recommends considering smoke damage to banks of interconnected panels, and this should be considered.
(This F&O originated from SR FSS-D9) 10-13         A credit for incipient detection is taken for    FSS-A6          This appears to be a documentation issue,   This F&O has been resolved.
MCB fires (non-suppression probability of       FSS-D7          but since the system is not yet installed,  The credit taken for incipient detection is 0.02). There is no documentation to justify      FSS-H7          there could be an impact to the assumptions consistent with that specified in FAQ                this value. Per discussion with FP&L the                         made within the Fire PRA.                   0046.
approach appears to be in agreement with                         Document the basis for probability of non-FAQ-08-0046. The approach also does not                         suppression value assumed in analysis.
use the NUREG/CR-6850 Appendix L factor                         When the incipient system is installed, the for panels that credit incipient detection.                      FPRA should be reviewed and updated Secondly, the incipient detection system is                      accordingly.
not yet installed, and therefore, the Fire PRA should be reviewed and updated as needed to reflect any differences between the Revision 0                                                                                                                                                      Page V-77

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR           Basis and Recommendation                       Disposition in Fire PRA Update assumed and as-built conditions of the system.
(This F&O originated from SR FSS-A6) 10-14          Beyond the Generic Fire Modeling                FSS-A5         For risk significant fire scenarios, detailed   This F&O has been resolved.
Treatments, the Fire PRA did not include                       fire modeling should be performed to ensure     The current analysis is consistent with a additional detailed fire modeling for most fire                you are not masking the "true risk significant  Capability Category I analysis. This compartments.                                                  fire areas". Without detailed fire modeling     provides a degree of conservatism in the Note 4 (under FSS-A5 of the ASME                               for significant fire scenarios, the results are analysis which would also tend to over-Standard) states that "once a fire scenario                    conservative,                                   estimate the change in risk which is reported has been 'selected,' this implies that the                     Consider performing additional detailed fire    for the NFPA 805 application. A review of scenario will eventually be evaluated and/or                   modeling to provide "reasonable assurance       the results of the application analyses quantified at a level of detail commensurate                   that the fire risk contribution of each         indicates more rigorous analyses consistent with the risk significance of the scenario."                   unscreened physical analysis unit can be        with CC II or CC III would not alter the (This F&O originated from SR FSS-A5)                           characterized."                                 conclusions of the analyses.
10-15          PTN credits multiple suppression paths for      FSS-C7         Lack of dependency analysis could lead to       This F&O is resolved.
MCA/HGL evaluation. However, the                FSS-G1         an optimistic estimate of suppression           A review of the credited suppression dependencies have not been evaluated and        FSS-H7         probability.                                    systems in the Multi-Compartment /Hot Gas modeled. For example, fixed suppression                        When multiple suppression paths are             Layer analysis has confirmed that no and fire brigade response may both rely on                     credited, perform a review and address any      dependency exists between the suppression a single detection system.                                     dependencies between suppression and            systems and detection systems. Detection (This F&O originated from SR FSS-C7)                           detection systems credited in the MCA/HGL       in the zones with suppression systems is calculation.                                    associated with an independent detection system.
10-16          Review of fire modeling in single               FSS-Cl         Discounting of secondary combustibles           This F&O has been resolved.
compartments does not consider the              FSS-D3         when considering localized fire damage          Supplemental walkdowns have been addition of HRR from secondary                  FSS-G1         could lead to non-conservative results.         performed to identify and address the combustibles. It is acknowledged that                          Include secondary combustibles in the heat      potential for fire spread for scenarios where secondary combustibles were considered for                     release rates used for zone of influence        the non-383 cables are not protected by the MCA/HGL evaluation.                                        estimates.                                      Flammastic material. The analysis has been Fire spread and additional HRR due to the                                                                      updated to include these scenarios as resulting cable tray fire and adjacent                                                                         appropriate.
cabinets would increase the total fire size and the subsequent zone of influence.
Compared to the NUREG/CR-6850 guidance for flame spread along PVC cable (flame spread = 0.9 mm/sec) the estimation Revision 0                                                                                                                                                          Page V-78 0                                                                            0

Florida Power & Light                                                                                                                Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                    Fire PRA SR              Basis and Recommendation                  Disposition in Fire PRA Update of HRR for the applied scenarios is non-conservative. Realistic estimation of the scenario HRR is necessary to ensure the full impact of the fire on exposed targets is presented and that the effects of a damaging HGL may also be estimated.
FP&L has stated that walkdowns are in progress to include fire spread to cable trays and incorporate this into the fire scenarios.
(This F&O originated from SR FSS-C1) 10-17          The 2010 peer review identified that           FSS-D1          Modifying the zone of influence to account   This F&O has been resolved.
Attachment B of the Fire Scenario Report                        for HGL effects could impact the defined     The MCA/HGL evaluation has been modified (Report 0493060006.004) generic fire                           target damage set.                           to address the concern noted in the F&O.
modeling treatments do not account for the                     The generic treatments used in relatively    The potential for a larger zone of influence is effects of hot gas layer (HGL) on the zones                    small rooms should be scrutinized to ensure  addressed via new fire scenarios added to of influence. The limitation indicates that                    that any HGL interaction is considered and   the fire PRA.
because HGL is not considered that these                       accounted for if found to be significant.
correlations should not be used in enclosed                    The selection of which generic fire modeling areas with small volumes where a significant                   treatment is used to define target damage HGL thickness may form. Because this                           for HGL effects on a scenario basis should relationship is not considered plume                           be documented in a clear manner to temperatures may be underestimated                              facilitate updates and peer reviews.
because it is assumed that ambient temperature air is being entrained into the plume, resulting in cooler plume temperatures, rather than heated air from the hot gas layer. Entrainment of heated air into the fire plume results in higher damage heights because the plume remains hotter at higher elevations.
Since this review, FP&L states that "The impact of a hot gas layer on the zone of influence is evaluated for all fire zones/scenarios in the MCA/HGL evaluation." A review of this evaluation confirms that HGL effects on ZOI were in fact considered for the generic treatments; however, there is not sufficient Revision 0                                                                                                                                                      Page V-79

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR           Basis and Recommendation                      Disposition in Fire PRA Update documentation in Attachment A to the Fire Scenario Report to determine which zone of influence was applied to which scenario, and whether it was applied correctly to consider the effects of HGL. The Generic treatments include several iterations and combinations of variables, including opening percentage of the compartment. The fire scenario documentation at the time of this review did not provide sufficient information on opening percentage to confirm that the ZOI was applicable to the compartment.
(This F&O originated from SR FSS-D1) 10-18          In at least two cases, transient fire scenarios FSS-A1         The exclusion of transients in some            This F&O has been resolved.
have not been included in the fire modeling                    compartments may lead to a non-                Supplemental walkdowns were performed to for some compartments (e.g., fire                              conservative estimate of CDF and LERF.         re-assess the treatment of transient fires.
compartments 67 and 68). Per discussion                                                                       These walkdowns focused on two key with FP&L the transients may have been                                                                        attributes - the appropriateness of the excluded based on the dominance of the                         Include transient scenarios in all compartments where fire modeling has been      selected HRR characterization and the frequency of fixed scenarios. However,                                                                        location of the postulated fire scenarios.
transients should only be excluded when                        employed.
The postulated location for the treatment of precluded by design. Based on the size of                                                                     transient fires was based on where a these rooms, and the presence of secondary                                                                    transient ignition source might reasonably combustibles, transient fires could lead to                                                                   occur. The results of these walkdowns were fire growth and eventually HGL, and                                                                           incorporated into the FPRA analysis.
therefore should be analyzed.
(This F&O originated from SR FSS-A1) 10-19          For fire modeling analysis of transient fires,  FSS-H1         Lack of documentation on transient fire        This F&O has been resolved.
FP&L implements a floor area weighting                         locations and boundaries will present a        The specific instance noted in the F&O was factor. However, the documentation does                        challenge for updates and peer reviews,        corrected. In addition, supplemental not include a graphical representation of the                  Update documentation to include a graphical    walkdowns were performed to re-assess the assumed transient locations and                                representation of transient fire locations and overall treatment of transient fires. These boundaries. It is therefore not possible to                    boundaries.                                    walkdowns focused on two key attributes -
review (or update) transient fires.                                                                           the appropriateness of the selected HRR Also during review of transient weighting                                                                     characterization and the location of the factors it appears to have been double                                                                        postulated fire scenarios. However, the counted in some compartments (e.g.,                                                                           documentation that was generated did not Revision 0                                                                                                                                                         Page V-80

Florida Power & Light                                                                                                                  Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR             Basis and Recommendation                    Disposition in Fire PRA Update compartment 63). Based on discussion with                                                                     specifically produce graphical FP&L this was due to an error in the Excel                                                                    representations. Instead, the information based spreadsheet tool for transient                                                                          was incrementally enhanced to provide a frequency quantification. This appears to be                                                                  spatial reference to a location with in the an isolated case and will be corrected.                                                                       space. The need for special depiction of (This F&O originated from SR FSS-H 1)                                                                         transient fire scenario locations will be addressed in conjunction with the development of procedures for post transition configuration control.
10-20          The fire modeling analysis of the Turbine      FSS-Al          Lack of consideration of the catastrophic T/G  This F&O has been resolved.
Generator (T/G) fires is performed in                          fire may lead to a non-conservative estimate   The analysis documentation has been accordance with Appendix 0 to NUREG/CR-                        of CDF and LERF.                               updated to address catastrophic T/G fires 6850. However, there is no discussion                          Perform a review of the catastrophic T/G fire  that may lead to building collapse or other regarding the lack of analysis of the                          in accordance with Appendix 0 to               significant widespread damage. Theresults catastrophic T/G fire event, which should                      NUREG/CR-6850, or document the                 of this update did not identify any new risk consider blade ejection, oil line rupture, and                 justification for excluding this event at PTN. significant contributors or insights.
hydrogen explosion. Per discussion with FP&L, the catastrophic fire was discounted since the T/G is located outdoors. While this may not result in hot gas layer formation and structural collapse, a review of the guidance is warranted, and inclusion of this event frequency should as a minimum map to the loss of the T/G and if suppression fails, all equipment within the T/G structure.
(This F&O originated from SR FSS-Al)
Page V-81 Revision 00 Revision                                                                                                                                                            Page V-81

Florida Power & Light                                                                                                                    Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                     Fire PRA SR           Basis and Recommendation                        Disposition in Fire PRA Update 10-21          The supplemental generic Fire Model             FSS-C3         The current approach results in many             This F&O has been resolved.
Treatments: Transient Ignition Source           FSS-G1         transient fire scenarios being screened          Supplemental walkdowns were performed to Strength includes an assumption for                            during detailed scenario analysis. The           re-assess the treatment of transient fires.
transient burnout of 12 minutes. This           FS-H2          results of the FPRA are therefore potentially    These walkdowns did not identify any burnout time is based on an assumed fire                       non-conservative for the analyzed detailed       instances where an altering of the transient loading and the 317kW heat release rate,                       scenarios,                                       fire duration had any material impact on the and appears to be optimistic given the                         Provide additional justification for the applied HGL and MCA. The documentation has uncertainty in transient fire loading. The                     transient fire analysis as a screening           also been updated to address the criteria burnout is then used to develop a zone of                      approach. Consider increasing the burnout        used for selecting the characteristic transient influence for thermoplastic targets, based on                  time and using the NUREG/CR-6850                 fire HRR. The approach is consistent with the thermal response tables in Appendix H                      recommended damage threshold to 205&deg;C            the recently issued guidance from the to NUREG/CR-6850 for thermoplastic cable                       to bound uncertainties in fuel loading for       EPRI/NRC review panel. The results of at 260*C. Since this resultant vertical zone                   transient fires.                                 these walkdowns were incorporated into the of influence is used to screen transient                                                                        FPRA analysis.
scenarios from impacting secondary targets                                                                      The twelve minute fire corresponds to the higher than 7.3 feet from the floor, additional                                                                 317 kWfuel package only and represents justification is needed to demonstrate that a                                                                    35 lb. of Class A material. Additional 12 minute fire, and subsequent use of 260oC                                                                     discussion is provided in Rev. 0 of damage threshold is appropriate for                                                                             Supplement 3 of the Hughes Generic Fire screening purposes.                                                                                             Modeling treatments that examines the fire Also noted is that Attachment B to the Fire                                                                      durations and test durations of all Scenario Report zone of influence does not                                                                      NUREG/CR 6850 tests. It is shown that the reflect the same values recommended by                                                                          method used to determine a 12 minute fire the Generic Fire Model Treatment. As an                                                                         predicts or overestimates the fire duration in example, the differentiation between                                                                            all cases and is therefore a sound approach.
transient Severe and Non-Severe categories is not based on a 317kW fire. This appears to be a documentation issue only.
(This F&O originated from SR FSS-C3)
Revision 0                                                                                                                                                           Page V-82 0

Florida Power & Light                                                                                                              Attachment V - Fire PRA Quality Table V-3 DISPOSITION OF 2010 PTN FIRE PRA PEER REVIEW 'FINDING' F&Os Finding F&O                     Discussion                   Fire PRA SR             Basis and Recommendation                  Disposition in Fire PRA Update 10-22          Per NUREG/CR-6850, appendix H,                FSS-C6          For smaller volume rooms, estimates of       This F&O has been resolved.
temperature sensitive equipment should be     FSS-G1          equipment damage may be non-                 The consideration of sensitive electronics considered to fail at 65*C. Supplemental                      conservative,                                was addressed in a qualitative fashion in the Generic Fire Model Treatments: Hot Gas                        Apply the appropriate hot gas layer and      Scenario Report.
Layer Tables includes new zone of influence                   zone of influence for temperature sensitive and hot gas layer treatments for temperature                  equipment where applicable.
sensitive equipment. However, per discussion with FP&L these have not been implemented in the fire scenarios.
(This F&O originated from SR FSS-C6) 10-23          The PTN FSS report 0493060006.004, Rev        FSS-C5          Equipment damaged by suppression             This F&O has been resolved.
2, section 6 discusses the damage criteria                    activities may impact estimates of CDF and   The specific issue raised in the F&O is for thermal, smoke, and sensitive                             LERF for some scenarios,                     beyond the scope of the associated SR. In equipment. However, suppression effects                       Perform an assessment of electrical          addition, no known consensus method do not appear to have been considered for                     equipment that may be vulnerable to water    exists for treatment. A qualitative the potential to damage equipment.                            intrusion from suppression activities (or    assessment, based on other guidance for (This F&O originated from SR FSS-C5)                          thermal shock from gaseous systems), and     evaluation of potential impact of suppression include any additional failed equipment, not effects was performed which indicated that already considered damaged by fire, in       no specific change in the analysis is needed.
scenarios as appropriate.
Revision 0                                                                                                                                                     Page V-83}}