ML13267A212

Enclosure 1, Transition Report - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition

ML13267A212
Person / Time
Site:	Robinson
Issue date:	09/30/2013
From:	Duke Energy Progress
To:	Office of Nuclear Reactor Regulation
References
RNP-PA/13-0090
Download: ML13267A212 (410)

Text

United States Nuclear Regulatory Commission Enclosure I to Serial: RNP-RAI13-0090 975 Pages with Cover Page ENCLOSURE 1 TRANSITION REPORT

Duke Energy H. B Robinson Steam Electric Plant Unit No. 2 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition SDUKE

.ENERGY, Transition Report September 30, 2013

Duke Energy NFPA 805 Transition Report TABLE OF CONTENTS Executive Summary .................................................................................................. iv Acronym List .......................................................................................................... v

1.0 INTRODUCTION

..................................................................................................... 1 1.1 B ackg 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 u rpo 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 .................................................................................... 8 3 .1 B ackg ro und .................................................................................................. .. 8 3.2 NFPA 805 Process ........................................................................................ 8 3.3 NEI 04 NFPA 805 Transition Process .................................................... 9 3.4 NFPA 805 Frequently Asked Questions (FAQs) .......................................... 10 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS ........................................ 12 4.1 Fundamental Fire Protection Program and Design Elements ...................... 12 4.1.1 Overview of Evaluation Process ........................................................ 12 4.1.2 Results of the Evaluation Process ...................................................... 14 4.1.3 Definition of Power Block and Plant .................................................... 15 4.2 Nuclear Safety Performance Criteria ........................................................... 15 4.2.1 Nuclear Safety Capability Assessment Methodology ........................... 15 4.2.2 Existing Engineering Equivalency Evaluation Transition .................... 22 4.2.3 Licensing Action Transition .................................................................. 23 4.2.4 Fire Area Transition ............................................................................. 24 4.3 Non-Power Operational Modes .................................................................... 27 4.3.1 Overview of Evaluation Process ........................................................ 27 4.3.2 Results of the Evaluation Process ...................................................... 30 4.4 Radioactive Release Performance Criteria .................................................. 30 4.4.1 Overview of Evaluation Process ........................................................ 31 4.4.2 Results of the Evaluation Process ...................................................... 31 4.5 Fire PRA and Performance-Based Approaches .......................................... 32 4.5.1 Fire PRA Development and Assessment ............................................. 32 HBRSEP LAR Rev 0 Page i

Duke Energy NFPA 805 Transition Report 4.5.2 Performance-Based Approaches ......................................................... 34 4.6 Monitoring Program .................................................................................... 39 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program ...... 39 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ............... 39 4.7 Program Documentation, Configuration Control, and Quality Assurance ........ 45 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 8 0 5 ..................................................................................................... . . 45 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of N FPA 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 Results ..................................................................................... 54 4.8.1 Results of the Fire Area Review ........................................................ 54 4.8.2 Plant Modifications and Items to be Completed During the Implementation P hase ................................................................................................ . . 54 4.8.3 Supplemental Information -Other Licensee Specific Issues ............... 55

5.0 REGULATORY EVALUATION

......................................................................... 63 5.1 Introduction - 10 CFR 50.48 ...................................................................... 63 5.2 Regulatory Topics ...................................................................................... 68 5.2.1 License Condition Changes ............................................................... 68 5.2.2 Technical Specifications ...................................................................... 68 5.2.3 Orders and Exemptions ...................................................................... 68 5.3 Regulatory Evaluations ............................................................................... 68 5.3.1 No Significant Hazards Consideration ................................................. 68 5.3.2 Environmental Consideration ............................................................. 69 5.4 Revision to the UFSAR .................................................................................. 69 5.5 Transition Implementation Schedule ........................................................... 69

6.0 REFERENCES

.................................................................................................. 70 ATTACHMENTS ....................................................................................................... 75 A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program &

Design Elements ........................................................................................... A-1 B. NEI 04-02 Table B Nuclear Safety Capability Assessment - Methodology Review ........................................................................................................... B-1 C. NEI 04-02 Table B Fire Area Transition .................................................. C-1 D. NEI 04-02 Non-Power Operational Modes Transition ................................. D-1 E. NEI 04-02 Radioactive Release Transition .................................................. E-1 Page ii HBRSEP LAR Rev 0 LAR Rev 0 Page ii

Duke Energy NFPA 806 Transition Report F. Fire-Induced Multiple Spurious Operations Resolution ............................. F-1 G. Recovery Actions Transition ....................................................................... G-1 H. NFPA 805 Frequently Asked Question Summary Table ............................ H-1 I. Definition of Power Block .............................................................................. I-1 J. Fire Modeling V&V ........................................................................................ J-1 K. Existing Licensing Action Transition ........................................................... K-1 L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))... L-1 M. License Condition Changes ......................................................................... M-1 N. Technical Specification Changes ............................................................... N-1

0. Orders and Exemptions ................................................................................ 0-1 P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4) ................................... P-1 Q. No Significant Hazards Evaluations ........................................................... Q-1 R. Environmental Considerations Evaluation ................................................. R-1 S. Modifications and Implementation Items .................................................... S-1 T. Clarification of Prior NRC Approvals ........................................................... T-1 U. Internal Events PRA Quality ......................................................................... U-1 V. Fire PRA Q uality ................................................................................................ V-1 W. Fire PRA Insights .................................................................................................. 1 Rev 00 LAR Rev Page iii HBRSEP HBRSEP LAR Page iii

Duke Energy Executive Summary Executive Summary Duke Energy will transition the H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP) fire protection program to a new Risk-Informed, Performance-Based (RI-PB) alternative per 10 CFR 50.48(c) which incorporates by reference NFPA 805. The licensing basis per License Condition 3.E will be superseded.

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

• Required by 10 CFR 50.48(c).

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

• Recommended by guidance document Regulatory Guide 1.205 Revision 1.

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

" Fundamental Fire Protection Program Elements and Minimum Design Requirements

" Nuclear Safety Performance Criteria, including:

o Non-Power Operational Modes o Fire Risk Evaluations

" Radioactive Release Performance Criteria

" Monitoring Program

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

" Changes to License Condition

" Changes to Technical Specifications, Orders, and Exemptions,

" Determination of No Significant Hazards and evaluation of Environmental Considerations.

The attachments to the Transition Report include detail to support the transition process and results.

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

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Duke Energy Acronym List Acronym List AC Alternating Current ADAMS Agency wide Documents Access and Management System AFW Auxiliary Feedwater AHJ Authority having jurisdiction ANS American Nuclear Society AO Auxiliary Operator APCSB Auxiliary Power Conversion Systems Branch ARCTM Safe Shutdown Analysis software package ASME American Society of Mechanical Engineers BNP Brunswick Nuclear Plant BTP Branch Technical Position CAFTA Computer Aided Fault Tree Analysis CAT Capability Category CC Capability Category CC I Capability Category I CCDP Conditional Core Damage Probability CCW Component Cooling Water CDF Core Damage Frequency CFAST Consolidated Model of Fire and Smoke Transport CFR Code of Federal Regulation CO 2 Carbon Dioxide CP&L Carolina Power and Light CR3 Crystal River Unit 3 Nuclear Power Plant CSDB Component Selection Database CST Condensate Storage Tank CT Current Transformer CV Containment Vessel CVCS Chemical and Volume Control System DBD Design Basis Document HBRSEP LAR Rev 0 Page v

Duke Energy Acronym List Duke Energy Acronym List DC Direct Current DID Defense-in-Depth DSDG Dedicated Shutdown Diesel Generator EC Engineering Change EDB Equipment Database EDG Emergency Diesel Generator EEE Engineering Equivalency Evaluations EEEE Existing Engineering Equivalency Evaluations EOOS Equipment Out of Service EPRI Electric Power Research Institute ESP Engineering Support Personnel F&O Facts and Observations FA Fire Area FAQ Frequently Asked Question FC Fire Compartment FDT Fire Dynamics Tools FHA Fire Hazards Analysis FHB Fuel Handling Building FMEA Failure Modes and Effects Analysis FP Fire Protection FPIP Fire Protection Initiatives Project FPP Fire Protection Program FPRA Fire Probabilistic Risk Analysis or Assessment FRE Fire Risk Evaluation FRN Federal Register Notice FSA Fire Safety Analysis FSS Fire Scenario Selection FSSPMD Fire Safe Shutdown Program Manager Database FTL Fault Tree Logic FZ Fire Zone GDC General Design Criterion Page vi HBRSEP LAR Rev 0 LAR Rev 0 Page vi

Duke Energy Acronym List Duke Energy Acronym List HBRSEP H. B. Robinson Steam Electric Plant Unit No. 2 (i.e., RNP)

HEAF High Energy Arcing Fault HNP Shearon Harris Nuclear Power Plant HRE Higher Risk Evolutions HSS High Safety Significance HVAC Heating, Ventilation and Air Conditioning INPO Institute of Nuclear Power Operations ISFSI Independent Spent Fuel Storage Installation KSF Key Safety Function kV Kilovolt LA Licensing Action LAR License Amendment Request LERF Large Early Release Frequency LFS Limiting Fire Scenario LOCA Loss of Coolant Accident LSS Low Safety Significance MCA Multi-Compartment Analysis MCC Motor Control Center MCR Main Control Room MDAFW Motor Driven Auxiliary Feedwater MEFS Maximum Expected Fire Scenario MHIF Multiple High Impedance Fault MSO Multiple Spurious Operation MTC Moderator Temperature Coefficient NEI Nuclear Energy Institute NFPA National Fire Protection Association NFPA 805 National Fire Protection Association Standard 805 NGG Nuclear Generation Group NPO Non-Power Operations NRC Nuclear Regulatory Commission NSCA Nuclear Safety Capability Assessment Page vii HBRSEPLARRevO HBRSEP LAR Rev 0 Page vii

Duke Energy Acronym List Dueneg AcoymLs NSEL Nuclear Safety Equipment List OL Operating License OMA Operator Manual Action OOS Out-of-Service PAP Personnel Access Point PB Performance Based PORV Power Operated Relief Valves POS Plant Operational State PRA Probabilistic Risk Assessment or Analysis PSA Probabilistic Safety Assessment or Analysis PVC Polyvinyl-chloride PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group PWST Primary Water Storage Tank QA Quality Assurance RA Recovery Action RAB Reactor Auxiliary Building RA-DID Recovery Action - Defense-in-Depth RAI Request for Additional Information RAW Risk Achievement Worth RCA Radiologically Controlled Area RCP Reactor Coolant Pump RCS Reactor Coolant System RG Regulatory Guide RHR Residual Heat Removal RI-PB Risk-Informed Performance-Based RIS Regulatory Issues Summary RMA Risk Mitigating Action RWST Refueling Water Storage Tank SBO Station Blackout SDAFW Steam Driven Auxiliary Feedwater Page viii HBRSEP LAR Rev 0 LAR Rev 0 Page viii

Duke Energy Acronym List Duke Energy Acronym List SE Safety Evaluation SER Safety Evaluation Report SFPE Society of Fire Protection Engineers SI Safety Injection SISBO Self-Induced Station Blackout SM Safety Margin SR Supporting Requirement SSA Safe Shutdown Analysis SSC Structures, Systems, and Components SSD Safe Shutdown SW Service Water UET Unfavorable Exposure Time UFSAR Updated Final Safety Analysis Report V&V Verification and Validation VCT Volume Control Tank VFDR Variances from the deterministic requirements ZOI Zone of Influence Page ix HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page ix

Duke Energy 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). Duke Energy 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 H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP) 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 HBRSEP 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.

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1.0 Introduction 1.0 IntroductIon Duke nerg Duke Energy

- - Incorporation by 50.48(c)

C - Reference 0

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. -. National Fire I*

• .... *di -l= r

• Protection Association Standard NFPA 805 NEI 04-02 RG 1.205 o* *' I Endorsement GUIDANCE FOR RI-PB FP FOR EXISTING IMPLEMENTING A RI-IPB LIGHT-WATER NUCLEAR FP PROGRAM UNDER 10 POWER PLANTS CFR 50.48(c)

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

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

The NRC expected approximately 23 LARs by the end of June 2011. As a result, the Commission worked with industry to develop and create a staggered LAR submittal schedule. On April 14, 2011, the NRC held a public meeting, during which the staff and stakeholders discussed the staggered approach method. In a letter (ML111101452)

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Duke Energy 1.0 Introduction dated April 20, 2011, the Commission approved the staffs recommendation to develop a staggered submittal and review process for these reviews, and submit a revision to the Enforcement Policy for Commission approval which would propose to extend enforcement discretion to correspond with the new LAR submittal dates. In a letter (ML11164A047) dated June 10, 2011, the Commission approved the staff's recommendation to publish the Federal Register Notice (FRN) announcing the revision to the Enforcement Policy to extend the enforcement discretion to correspond with a staggered LAR submittal schedule. On June 29, 2011, Progress Energy submitted a letter (MLI 11188A058) requesting extension of their enforcement discretion and committed to the submittal date of September 30, 2013.

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

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

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

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

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

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

2) Summarize the results of the transition process

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

4) Describe the new fire protection licensing basis

5) Describe the configuration management processes used to manage post-transition changes to the station and the fire protection program, and resulting impact on the licensing basis Page 3 HBRSEP HBRSEP LAR Rev 0 LAR Rev 0 Page 3

Duke Energy 2.0 Overview of Existing Fire Protection Program 2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM 2.1 Current Fire Protection Licensing Basis H. B. Robinson Steam Electric Plant, Unit No. 2 was licensed to operate on July 31, 1970, with a Renewed Facility Operating License, dated June 27, 2007. As a result, the HBRSEP fire protection program is based on compliance with 10 CFR 50.48(a), 10 CFR 50.48(b), and the following License Condition:

Duke Energy HBRSEP License Condition 3.E states:

E. Fire Protection Program Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Report dated February 28, 1978, and supplements thereto. Carolina Power & Light Company may make changes to the approved Fire Protection Program without prior approval 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, CP&L performed a fire hazards analysis which analyzed the HBRSEP fire protection program against the guidance of Appendix A to Branch Technical Position (BTP) Auxiliary Power Conversion Systems Branch (APCSB) 9.5-1. CP&L submitted the HBRSEP Fire Hazards Analysis (FHA) and response to Appendix A on December 29, 1976. Subsequent to the submittal of the FHA, additional NRC review of the Fire Protection Program (FPP) took place in the form of written staff questions and CP&L responses, meetings and telephone conferences with the staff. The NRC accepted certain aspects of the program, while CP&L committed to make changes to other portions of the program. The final acceptance of the HBRSEP Fire Protection Program was documented in NRC SER dated February 28, 1978. Open issues from this SER were documented in supplements to the SER dated September 4, 1979, February 21, 1980, and December 8, 1980.

The NRC forwarded Section 10 CFR 50.48 and Appendix R to 10 CFR 50 to CP&L on November 24, 1980. This regulation became effective on February 17, 1981. Since HBRSEP was licensed prior to January 1, 1979, it was required to meet only certain provisions of Appendix R. CP&L sought and received such an exemption from the Appendix R requirements.

Supplemental SER for Sections III.G and II1.L was issued on November 21, 1985 resolving the open items concerning spurious operations of high-low pressure interface valves due to a postulated fire (hot short). Exemptions from certain requirements of Section III.G.2 and III.G.3 were also granted by the NRC in letters dated November 13, 1981, November 25, 1983, October 25, 1984, September 17, 1986, June 30, 1988, and October 17, 1990.

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Duke Energy 2.0 Overview of Existing Fire Protection Program In addition to the approval of HBRSEP alternative shutdown design, the NRC granted the following:

02/28/78 NLU-78-71 Amendment 31 to Operating License (OL) adds license conditions, revises Technical Specifications and issues Appendix A to BTP APCSB 9.5-1 Fire Protection SER.

09/04/79 NLU-79-398 Amendment 40 to OL, changing Fire Protection license conditions and Suppl. 1 to Appendix A Fire Protection SER extending completion dates for modifications.

02/21/80 NLU-80-106 NRC evaluations of issues relating to Appendix A Fire Protection SER, identifying completed, open and not acceptable issues.

12/08/80 NLU-80-623 SER closing several Fire Protection issues related to Appendix A Fire Protection SER.

05/15/81 NLU-81-245 Amendment 57 to OL, revising Technical Specifications.

11/13/81 NLU-81-564 NRC grants exemption to Appendix R Section III.G.3 for a fixed fire suppression system in the control room.

05/10/82 NLU-82-248 NRC grants exemption to 10 CFR 50.48c schedular requirements with criteria for evaluation.

12/10/82 NLU-82-709 NRC Letter Regarding DRAFT SER on Appendix R Exemption Request for Sections III.G.2, III.G.3, Ill.L, HII.M and 111.0 11/22/83 NLU-83-772 SER related to Appendix R Sections III.G.3 and III.L.

Also denies CP&L request for exemption to achieve cold shutdown using onsite power within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Superceded by 8/8/84 SER.

11/25/83 NLU-83-777 NRC grants exemption to Appendix R Section III.G.2 requirements for FZ 27 for 3-hour fire barrier separation between redundant trains and FZ 29 for automatic suppression and 20 foot separation. NRC also grants exemption to Appendix R Section III.M.2 for fire barrier penetration seal backface temperatures and approval of seals having a 2-hour fire rating.

05/21/84 NLS-84-155 Informal forwarding of DRAFT Supplemental SER on III.G.3 and III.L.

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Duke Energy 2.0 Overview of Existing Fire Protection Program Duke Energy 2.0 Overview of Existing Fire Protection Program 08/08/84 NLU-84-516 SER for Appendix R Sections III.G.3 and III.L.

Supersedes 11/22/83 SER (sometimes referred to as SSER or Revised SER).

10/25/84 NLU-84-687 NRC grants exemption to Appendix R Section III.G.2 for FZ 5 from area wide automatic fire suppression.

03/07/85 NLS-85-146 NRC Letter Regarding Exemption To Appendix R Section 111.0, Notice of Environmental Assessment and Finding of No Significant Impact 03/07/85 NLU-85-176 NRC grants exemption to Appendix R Section 111.0 from installation of reactor coolant pump oil collection system.

11/21/85 NLS-85-732 Supplemental SER for Appendix R Sections III.G. and III.L resolving open item concerning spurious operation of the high-low pressure interface valves due to a postulated fire (hot short).

09/09/86 NLS-86-552 NRC Letter Regarding Additional Information Regarding Exemption From Certain Requirements of 10 CFR Part 50, Appendix R, Sections III.G.2 and III.G.3 09/11/86 NLS-86-551 NRC Letter Regarding Environmental Assessment on Exemption Request from Certain Requirements of 10 CFR Part 50, Appendix R, Sections IIl.G.2.f and III.G.3 09/17/86 NLU-86-570 NRC Letter Granting Exemption from Certain Requirements of 10 CFR Part 50, Appendix R, Sections III.G.2.f for radiant heat shield in FZ 24 and III.G.3 for partial fire detection and suppression in Fire Areas A (FZs 3, 6, 7, 8, 11, 12, 13, 15, 16, 17, 18, 21, 23),

B (FZ 4), and G (FZs 25, 28, 30, 31, 32, 33) 07/30/87 NLS-87-422 NRC Letter Regarding Exemptions from Certain Requirements of 10 CFR Part 50, Appendix R, Section IIl.J. Superseded by June 30, 1988.

06/30/88 NRC-88-390 NRC Letter Granting Exemptions From Certain Requirements of 10 CFR Part 50, Appendix R, Section III.J for cold shutdown equipment areas, along alternate egress routes outside buildings and use of dedicated portable hand held lighting. Supersedes July 30, 1987 Exemption due to inconsistencies between CP&L letters and NRC SER.

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Duke Energy 2.0 Overview of Existing Fire Protection Program Duke Energy 2.0 Overview of Existing Fire Protection Program 10/09/90 NRC-90-600 NRC Letter Regarding Environmental Assessment For Appendix R III.G.2 Exemption 10/17/90 NRC-90-622 NRC Letter Granting Exemption From Requirements of Section III.G.2.b of Appendix R for increased combustible loading up to a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in CCW Pump Room (FZ 5) 10/02/92 NRC-92-563 NRC Letter Regarding Environmental Assessment Related to Exemption From Section III.J of Appendix R 10/08/92 NRC-92-581 NRC Letter Granting Exemption From The Requirements of 10 CFR Part 50, Appendix R Section IIl.J for cold shutdown equipment areas and use of dedicated portable hand held lighting. SER also clarifies 6/30/88 Exemption allowing use of portable lights at the intake structure (FZ 29).

12/07/92 NRC NRC Issues Amendment No. 142 to Facility Operating 0702 License No. DPR-23 Regarding Fire Protection - Revising Fire Protection License Condition of Operating License and relocates the Fire Protection Tech Specs to plant procedures and UFSAR 02/09/96 NRC Environmental Assessment and Finding of No Significant 0080 Impact Regarding and Exemption from Requirements of 10CFR Part 50, Appendix R, Section IIl.J 06/03/96 NRC Exemption From Certain Requirements of 0235 10CFR Part 50, Appendix R, Section III.J for access to and egress from, and operation of valves in outside areas which are illuminated by diesel backed security lighting.

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Duke Energy 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, Duke Energy 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.

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Duke Energy 3.0 Transition Process Establish fundamental fire NFPA 805 Section 2.2(a) protection elements (Chapter 3)

Identify fire hazards NFPA805 Section 2.2(b)

Identify performance ri oe Nuclear safety c riteria to be Life safety NFPA 805 Section 2.2(c) examined Property damagetbusiness (Chapter 1) Interruption Evaluate compliance to I* Radiation release performance criteria Identify structures, systems, or componentsaire (SSCs) in each NFPA 805 Section 2.2(d) area to which the performance criteria applies NFPA 805 SecUon 2.2(e) 4,l Perfonmance-Besed Approach DeterminisatcApproach Evaluate ability to satisfy performance Maintain compliance with existing plant requirements license basis (10 CFR 50 App. R, Approved Exemptions. Engineering Evaluations) (Chapter 4)

S Existing PerformanceBasis DeterministicBasis Engineenng Define fire scenarios and fire design basis Verify deterministic requirements are met Equivalency Evaluations for each fire area being considered.

Evaluate using, e.g..

Fire modeling to quantify the fire risk NFPA 805 Section 2.2(f) and margin of safety PSA to examine impact on overall plant risk NFPA 806 Section 2.2(g)

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

Evaluate risk impact of changes to the approved design basis acepabe Examples Yes Design Basis Documents DocuenttionandconfgurtionI /Fire hazards analysis NFPA 805 Section 2.26) Documentation and configuration Nuclear safety capability assessment

'Ontrol Supportng engineeanng calculations

~Probabilistic safety analysis Risk-informed change evaluations Feedback Establish monitoring p NFPA 805 Section 2.2(l)

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.

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Duke Energy 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

IFP Fundamentals Review and Confirmation Nuclear Safety Review and Confirmation 1

[Identifyoutliers / VFDRs Transition Report Sect. 4.4 Transition Report Sect. 4.3 Fidentify outliers / VFDRs1 1 A - - II i

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} Sect. 4.5 Transition Verify / Establish Monitoring Report Program

} Sect. 4.6 Transition Transition Confirm / Establish Adequate Documentation I Quality and Report Configuration Control Sect. 4.7, 5 Regulatory Submittal and Approval

}eTransition Report 4.8,5

.Sec 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 HBRSEP LAR Rev 0 Page 10

Duke Enerqy 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 11 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page 11

Duke Energy 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 HBRSEP Fire Protection Program to the requirements of NFPA 805 Chapter 3 was performed and documented in Attachment A, Table B-I, NFPA 805 Ch. 3 Transition Details. The analysis used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 (See Figure 4-1).

Each section and subsection of NFPA 805 Chapter 3 was reviewed against the current fire protection program. Upon completion of the activities associated with the review, the following compliance statement(s) was used:

" Complies - For those sections/subsections determined to meet the specific requirements of NFPA 805

• Complies with Clarification - For those sections/subsections determined to meet the requirements of NFPA 805 with clarification

" Complies 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 with use of Existing Engineering Equivalency Evaluations (EEEEs) -

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

" 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 12 HBRSEPLARRevO HBRSEP LAR Rev 0 Page 12

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[Based on NEI 04-02 Figure 4-2]3 Figure 3 4-1 depicts the process used during the transition and therefore contains elements (i.e., open items) that represent HBRSEP LAR Rev interim resolutions. Additional detail on the transition of EEEEs 0 is included in Section 4.2.2.

Page 13

Duke Energy 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-1, 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 HBRSEP 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 NFPA 805 Section 3.1 states in part, "Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein." In some cases prior NRC approval of an NFPA 805 Chapter 3 program attribute may be unclear. Duke Energy does not have any requests for the NRC to concur with their finding of prior approval for any sections of NFPA 805 Chapter 3.

4.1.2.3 NFPA 805 Chapter 3 Requirements Not Met and Not Previously Approved by NRC The following sections of NFPA 805 Chapter 3 are not specifically met nor do previous NRC approvals of alternatives exist:

3.3.5.1- Approval is requested for the use electrical wiring above suspended ceilings.

3.3.5.2- Approval is requested for the use of electrical raceway construction that may not comply.

3.5.16- Approval is requested for the use of fire protection water for specific plant evolutions.

3.2.3(1)- Approval is requested for the use of EPRI Technical Report TR-1 006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program.

The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements are provided in Attachment L. Duke Energy requests NRC approval of these performance-based methods.

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Duke Energy 4.0 Compliance with NFPA 806 Requirements 4.1.3 Definition of Power Block and Plant Where used in NFPA 805 Chapter 3 the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility's pre-transition licensing basis.

The HBRSEP FHA assigned fire protection properties to defined Fire Areas and Fire Zones that are important for the safe shutdown of the plant. The Fire PRA analysis fire compartments are equivalent to defined Fire Zones in the plant. Therefore, detection systems, suppression systems and floor areas for defined Fire Zones are the same for Fire PRA fire compartments. The current Appendix R series drawings and the Fire Hazards analysis were utilized in the partitioning determination. For Fire Compartments FC400 through FC500 there are no current Fire Areas or Zones defined. These areas are currently considered to be part of the yard.

These structures are listed in Attachment I and define the "power block" and "plant".

4.2 Nuclear Safety Performance Criteria The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805.

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

4.2.1 Nuclear Safety Capability Assessment Methodology The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes:

• Establishing compliance with NFPA 805 Section 2.4.2

• Establishing the Safe and Stable Conditions for the Plant

" Establishing Recovery Actions

• Evaluating Multiple Spurious Operations The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3.

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Duke Energy 4.0 Compliance with NFPA 806 Requirements 4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states:

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

(1) Selection of systems and equipment and their 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 nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety 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 (ML050310295) Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. The methodology is depicted in Figure 4-2 and consisted of the following activities:

Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station.

The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section:

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

The comparison of the HBRSEP existing post-fire SSA methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in Attachment B.

In addition, a review of NEI 00-01, Revision 2, (ML091770265) Chapter 3, was conducted to identify the substantive changes from NEI 00-01, Revision 1 that are applicable to an NFPA 805 fire protection program. This review was performed and documented in Attachment B.

HBRSEP LAR Rev 0 Page 16

Duke Energy 4.0 Compliance with NFPA 805 Requirements Results from Evaluation Process The method used to perform the existing post-fire SSA with respect to selection of systems and equipment, selection of cables, and identification of the location of equipment and cables, either meets the NRC endorsed guidance from NEI 00-01, Revision 1, Chapter 3 (as supplemented by the gap analysis) directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B.

StePI Stop2 Step3 4

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

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

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

A review of the NSCA results indicated that there are defense-in-depth recovery actions that require manual operation of a rising stem valve in the fire area of concern. There are no ignition sources or in situ combustibles that would affect these valves, and operation of the valves would not be required for about two hours.

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

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

HBRSEP LAR Rev 0 Page 17

Duke Energy 4.0 Compliance with NFPA 805 Requirements 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 requirementsfor hot shutdown for a boiling water reactorand hot standby for a pressurized water reactor.Forall other configurations,safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling."

The nuclear safety goal of NFPA 805 requires "...reasonableassurance that a fire during any operationalmode and plant configuration will not prevent the plant from achieving and maintainingthe fuel in a safe and stable condition"without a specific reference to a mission time or event coping duration.

For the plant to be in a safe and stable condition, it may not be necessary to perform a transition to cold shutdown as currently required under 10 CFR 50, Appendix R.

Therefore, the unit may remain at or below the temperature defined by a hot standby/hot shutdown plant operating state for the event.

Results Based on the criteria discussed in NSCA calculation RNP-E/ELEC-1216, "The Fire Safe Shutdown Analysis for H.B. Robinson Nuclear Plant," the NFPA 805 licensing basis for HBRSEP is to achieve and maintain hot shutdown conditions following any fire occurring prior to establishing cold shutdown. Specifically, the conditions include:

" the reactor operating at power,

" a shutdown condition prior to aligning the RHR system for shutdown cooling, or

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

Immediately following the reactor scram, RCS inventory and pressure control is maintained using the charging system, or the safety injection system if the charging pumps are not available (applies to one fire area). Pressurizer safety relief valves provide overpressure protection for the RCS. Main Steam line safety relief valves provide for initial decay heat removal. Cycling of the steam generator power-operated relief valves (PORVs) provide for continued decay heat removal, with steam generator inventory replenished by either the steam driven auxiliary feedwater (AFW) pump or the motor-driven AFW pumps. When the CST is depleted, the suction of the AFW pumps is manually aligned to the service water system. These actions are required in about two hours, and are considered defense in depth recovery actions. Since the valves requiring operation are all manual valves and are not electrically supervised, they are not considered to be VFDRs.

HBRSEP LAR Rev 0 Page 18

Duke Energy 4.0 Compliance with NFPA 805 Reaulrements Following stabilization at hot standby, a long term strategy for decay heat removal and inventory/pressure control would be determined based on the extent of equipment damage. If an assessment of the post-fire conditions indicated that the residual heat removal (RHR) system should be in operation, then activities would commence in a safe and controlled manner to align plant equipment required for reactor cooldown.

The long-term actions required to maintain safe and stable conditions are largely routine and within the normal capabilities of site personnel, even in the face of fire damage.

These include the previously mentioned actions to align the suction of the AFW pump(s) to the service water system, and opening CVC-358 to maintain the charging pump suction path to the RWST. LCV-1 15B will initially provide this suction path, but is conservatively assumed to fail closed after about four (4) hours. These are straightforward actions performed by operators and covered by plant procedures. Repairs to safe shutdown equipment would not be required and the management of the onsite inventories of makeup water, nitrogen and diesel fuel would not require resources beyond those available from normal operations staff, maintenance, and emergency response personnel.

Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses.

" At-Power analysis, Modes 1-4. This analysis is discussed in Section 4.2.4. The at-power analysis goes beyond safe and stable to include Mode 4.

• Non-Power Operations analysis that includes cold shutdown and below (Modes 5 and 6). 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 2.2.1 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02 suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee's methodology.

The discussion below provides the methodology used to transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML110070485) and consists of the following steps:

" Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition.

• Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).

HBRSEP LAR Rev 0 Page 19

Duke Energy 4.0 Compliance with NFPA $05 Requirements

• 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 RNP-01 70 and RNP-0202. Refer to Attachment G for the detailed evaluation process and summary of the results from the process.

4.2.1.4 Evaluation of Multiple Spurious Operations Overview of Process NEI 04-02 suggests that a licensee submit a summary of its approach for addressing potential fire-induced MSOs for NRC review and approval. As a minimum, NEI 04-02 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 HBRSEP is summarized below.

As part of the NFPA 805 transition project, a review and evaluation of HBRSEP 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 PWR Generic MSO list from NEI 00-01, Revision 3 was utilized.

The approach outlined in Figure 4-3 (based on Figure XX 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, Rev. 1 Section F.4.2).

" Updating the Fire PRA model and existing post-fire NSCA to include the MSOs of concern.

• Evaluating for NFPA 805 Compliance.

" Documenting Results.

This process is intended to support the transition to a new licensing basis.

Post-transition changes would use the RI-PB change process. The post-transition change process for the assessment of a specific MSO would be a simplified version of this process, and may not need the level of detail shown in the following section (e.g.,

An expert panel may not be necessary to identify and assess a new potential MSO.

Identification of new potential MSOs may be part of the plant change review process and/or inspection process).

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Duke Energy 4.0 Compliance with NFPA 805 Requirements Identify Potential MSOs of Concern l SSA Step 1 l Generic List of MSOs

• 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 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.

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

• The EEEE is not based solely on quantitative risk evaluations,

• The EEEE is an appropriate use of an engineering equivalency evaluation,

" The EEEE is of appropriate quality,

" The standard license condition is met,

• The EEEE is technically adequate,

" The EEEE reflects the plant as-built condition, and

" The basis for acceptability of the EEEE remains valid In accordance with the guidance in RG 1.205, Regulatory Position 2.3.2 and NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are summarized in the LAR as follows:

If not requesting specific approval for "adequate for the hazard" EEEEs, then the EEEE was referenced where required and a brief description of the evaluated condition was provided.

If requesting specific NRC approval for "adequate for the hazard" EEEEs, then EEEE was referenced where required to demonstrate compliance and was included in Attachment L for NRC review and approval.

In all cases, the reliance on EEEEs to demonstrate compliance with NFPA 805 requirements was documented in the LAR.

Results The review results for EEEEs are documented in Attachment A. In all cases the EEEEs reflect the plant as-built condition and the basis for acceptability of the evaluation remains the same.

In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in the Attachments A and C as appropriate.

None of the transitioning EEEEs require NRC approval.

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Duke Energy 4.0 Compliance with NFPA 806 Requirements 4.2.3 Licensing Action Transition Overview of Evaluation Process The existing licensing actions (i.e., Appendix R exemptions) review was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following:

• Determination of the bases for acceptability of the licensing action.

• Determination that these bases for acceptability are still valid and required for NFPA 805.

• Additionally, variances from the deterministic requirements were identified in the NEI 04-02, Table B-3 (See Attachment C). Some of these variances were subsequently dispositioned via the use of the performance-based approach.

Results Attachment K contains the detailed results of the Licensing Action Review.

The licensing action review resulted in the identification of licensing actions that would be transitioned to the new 10 CFR 50.48(c) licensing basis and those that would no longer be necessary. Attachment K of the Transition Report contains the results of the Licensing Action Review.

Since the exemptions are either compliant with 10 CFR 50.48(c) or no longer necessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), CP&L requests that the exemptions listed in Attachment K be rescinded as part of the LAR process. It is CP&L's understanding that implicit in the superseding of the current license condition, all prior fire protection program Safety Evaluations and commitments will be superseded in their entirety.

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Duke Energy 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 HBRSEP LAR Rev 0 Page 24

Duke Energy 4.0 Compliance with NFPA 805 Requirements protection systems and features are subject to the applicable requirements of NFPA 805 Chapter 3.

Document Final Disposition of VFDR Compliance options include:

Accept As Is

- Require FP systemstfeatures Require Recovery Action Require Programmatic Enhancements Require Plant Modifications (B-3 Table)

Figure 4 Summary of Fire Area Review

[Based on FAQ 07-0054 Revision 1]

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Duke Energy 4.0 Compliance with NFPA 806 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 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 - HBRSEP is transitioning one Licensing Action. This exemption will remain part of the post-transition licensing basis. The exemption from Section 111.0 of Appendix R was granted by the NRC to the extent that a reactor coolant pump lube oil collection system is not provided. In lieu of installing such a system, fixed fire suppression is maintained and additional detection and dikes were installed in the pump bays. Also, the Containment Spray system serves as a backup fire suppression system with Sodium Hydroxide isolated. This is further explained in Attachment K.

" EEEE - Specific references to EEEE that rely on determinations of "adequate for the hazard" that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability should be provided.

• VFDRs - Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3. Refer to Section 4.5.2 for a discussion of the performance-based approach.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process HBRSEP implemented the process outlined in NEI 04-02 and FAQ 07-0040, Clarification on Non-Power Operations. 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.

Performed Fire Area Assessments (identify pinch points - plant locations where a single fire may damage all success paths of a KSF).

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.

Page 27 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page 27

Duke Energy 4.0 Compliance with NFPA 805 Requirements Duke Energy 4.0 Compliance with NFPA 805 RequIrements Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points Page 28 HBRSEP LAR Rev 0 LAR Rev 0 Page 28

Duke Energy 4.0 Compliance with NFPA 806 Requirements Duke Energy 4.0 Compliance with NFPA 805 RequIrements Higher Risk Evolution as Defined by Plant Specific Outage Risk Criteria for example

1) Time to Boil

2) Reactor Coolant System and Fuel Pool Inventory

3) Decay Heat Removal Figure 4-6 Manage Pinch Points Page 29 HBRSEP LAR Rev 0 LAR Rev 0 Page 29

Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.3.2 Results of the Evaluation Process HBRSEP outage management processes were reviewed. Based on FAQ 07-0040, the Plant Operating States considered for equipment and cable selection are documented in calculation RNP-E/ELEC-1217, "Non-Power Operations Analysis". Using a CAFTA fault tree that models NPO requirements, systems and components were identified to provide the following KSFs: Decay Heat Removal; Reactivity Control; Inventory Control; Pressure Control; Spent Fuel Pit Cooling; and Electrical Power Availability (to the extent that it supports the other KSFs).

For those components not already in the HBRSEP Access Database or those with a functional state for non-power operations differing from that in the At-Power Analysis, circuit analysis, cable selection and routing were performed as described in the plant's NSCA methodology. Once all information had been entered into the HBRSEP Access Database, the ARCTM software package in conjunction with the NPO fault tree was used to determine KSF Pinch Points.

Calculation RNP-E/ELEC-1217 provides the results of the fire area assessments for the Pinch Point analysis and provides recommendations for changes to fire risk and outage management procedures and other administrative controls. These include:

• Prohibition or limitation of hot work in fire areas during periods of increased vulnerability.

• Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability.

" Provision of additional fire watches in affected fire areas during increased vulnerability.

" Identification and monitoring of in-situ ignition sources for fire precursors (e.g.,

equipment temperatures).

• Review of work activities for possible rescheduling.

" Equipment realignment (e.g., swing pumps or Backfeed)

" Identification of procedures to be briefed or walked down.

• Posting of protected equipment.

• Consideration of pre-emptive or recovery actions to mitigate potential losses of KSF success paths.

Attachment D provides a more detailed discussion. Based on incorporation of the recommendations from RNP-E/ELEC-1217 into appropriate plant procedures in conjunction with establishment of the NFPA 805 fire protection program, the performance goal for NPO modes (i.e., to maintain the fuel in a safe and stable condition) is fulfilled and the requirements of NFPA 805 are met.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.4 Radioactive Release Performance Criteria 4.41 Overview of Evaluation Process The review of the fire protection program against NFPA 805 requirements for fire suppression related radioactive release was performed using the methodology contained in NEI 04-02, Table E-1, and was performed using the methodology contained in Project Instruction FPIP-0121, Radiological Release Reviews During Fire Fighting Operations, Rev. 1. The methodology consisted of the following:

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

The review panel consisted of representatives from Operations, Engineering (i.e.,

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

A review of engineering controls to ensure containment of gaseous and liquid effluents (i.e., smoke and fire fighting agents). This review included all plant operating modes (i.e., including full power and non-power conditions). Otherwise, provided a bounding analysis, quantitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of radioactive effluents specified in the unit's Technical Specifications are met.

4.4.2 Results of the Evaluation Process The review determined the Fire Protection Program (i.e., Pre-Fire Plans) meets the, radioactive release performance criteria by ensuring that radioactive materials (i.e.,

radiation) generated as a direct result of fire suppression activities is contained and monitored prior to release to unrestricted areas, such that release would be as low as reasonably achievable and would not exceed applicable 10 CFR, Part 20 limits.

Containment and monitoring is ensured through elements of the fire brigade training, guidance provided in pre-fire plans and certain plant features (i.e., engineering controls) such as curbs and ventilation systems or actions provided to control smoke management or fire suppression water run-off.

Site specific review of associated fire event and fire suppression related radioactive release is summarized in Attachment E, NEI 04-02, Table E-1. Containment and HBRSEP LAR Rev 0 Page 31

Duke Energy 4.0 Compliance with NFPA 805 Requirements monitoring actions associated with firefighting operations are included in the pre-fire plans for fire areas as appropriate based on the screening criteria previously stated (Attachment E) to meet the radiological performance criteria.

The standardized pre-fire plan outline identifies typical fixed radiological hazards for each area. All HBRSEP pre-fire plans were screened for applicability. Pre-fire plans that address areas where there is no possibility of radiological hazards were screened out from further review. This information was included as input to the individual fire area Fire Safety Analyses (FSAs) calculations. The FSA is the Design Basis Document for NFPA 805 compliance for each fire area and will serve as the location for maintenance and configuration control of the radioactive release review results. Change, modification, or revision to the FSAs is controlled under existing plant engineering configuration control processes.

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 HBRSEP in compliance with the requirements of Part 4 "Requirements for Fires At Power PRA," of the ASME and ANS combined PRA Standard, ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application," (hereafter referred to as Fire PRA Standard). Duke Energy conducted a peer review by independent industry analysts in accordance with RG 1.200 prior to a risk-informed submittal. The resulting fire risk assessment model is used as the analytical tool to perform Fire Risk Evaluations during the transition process.

Section 4.5.1.1 describes the Internal Events PRA model. Section 4.5.1.2 describes the Fire PRA model. Section 4.5.1.3 describes the results and resolution of the peer review of the Fire PRA, and Section 4.5.1.4 describes insights gained from the Fire PRA.

4.5.1.1 Internal Events PRA The HBRSEP base internal events PRA, Calculation RNP-F/PSA-0006 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. Attachment U demonstrates that the internal events PRA is met at Capability Category II for all applicable supporting requirements according to peer review and/or disposition.

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 Page 32 HORSEPLARRevO HBRSEP LAR Rev 0 Page 32

Duke Energy 4.0 Compliance with NFPA 805 Requirements Fire PRA was developed using the guidance for Fire PRA development in NUREG/CR-6850/EPRI TR 1011989, approved FAQs, and EPRI TR 1016735.

The Fire PRA quality and results are discussed in the subsequent sections and in Attachments V and W, respectively.

Fire Model Utilization in the Application As part of the NFPA 805 transition, fire modeling was performed as part of the Fire PRA development (i.e., NFPA 805 Section 4.2.4.2) and, therefore, maximum expected fire scenario (MEFS)/limiting fire scenario (LFS) were not analyzed separately. RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis.

The following fire models were used:

" Fire Dynamics Tools (FDT's)

" Consolidated Model of Fire and Smoke Transport (CFAST)

The approach taken at HBRSEP to simplify the analysis process incorporates features of several fire model tools covered by NUREG-1824, as well as additional features. The approach is collectively referred to as the Fire Modeling Generic Treatments. The analysis basis and Verification and Validation (V&V) documentation was provided in a proprietary Hughes Associates, Inc. report to the NRC on January 24, 2008. The report entitled "Generic Fire Modeling Treatments" is effectively a technical reference guide, a user's guide, and the V&V basis.

The use of the Generic Treatments in specific applications at HBRSEP falls within their limitations as described in the "Generic Fire Modeling Treatments". In addition to the generic fire modeling treatments that were used in the hazard analysis, several calculations were produced that used CFAST and the FDT's as documented in NUREG-1824.

The acceptability of the use of these fire models is included in Attachment J.

4.5.1.3 Results of Fire PRA Peer Review The HBRSEP Fire PRA Calculation RNP-F/PSA-0094 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 Fire PRA peer review report (March 2013) and subsequent focused-scope peer-review report (July 2013) were used to support the Fire PRA for the NFPA 805 application.

The Fire PRA update addressed the Supporting Requirement assessed deficiencies (i.e., Not Met or Capability Category I (CC I)). Completion of recommendations related to Supporting Requirement assessments and 'Finding' F&Os results in a Capability Category II assessment for the associated Supporting Requirements. Any Supporting Requirements found not to meet Category II are considered "Open," but were justified to HBRSEP LAR Rev 0 Page 33

Duke Energy 4.0 Compliance with NFPA 805 Requirements have an insignificant impact on the NFPA 805 application through disposition. The results of the peer review and dispositions are summarized in Attachment V.

4.5.1.4 Risk Insights Risk insights were documented as part of the development of the Fire PRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for fire PRA development and is useful in identifying the areas of the plant where fire risk is greatest.

A review of the fire scenarios contributing more than 1% to the overall 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).

• 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 demonstrating compliance with NFPA 805 for HBRSEP.

4.5.2.2 Fire Risk Approach Overview of Evaluation Process The Fire Risk Evaluations were completed as part of the HBRSEP NFPA 805 transition.

These Fire Risk Evaluations were developed using the process described below. This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These are summarized in Table 4-1.

Table 4-1 Fire Risk Evaluation Guidance Summary Table Document Section(s) Topic NFPA 805 2.2(h), 4.2.4, A.2.2(h), A.2.4.4, D.5 Change Evaluation (2.2(h), 2.2.9, 2.4.4 A.2.2(h), A.2.4.4, D.5)

Risk of Recovery Actions (4.2.4)

Use of Fire Risk Evaluation (4.2.4.2)

NEI 04-02 Revision 2 4.4, 5.3, Appendix B, Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (App. I), No specific discussion of Fire Risk Evaluation RG 1.205 Revision 1 C.2.2.4, C.2.4, C.3.2 Risk Evaluations (C.2.2.4)

Recovery Actions (C.2.4)

During the transition to NFPA 805, variances from the deterministic approach in Section 4.2.3 of NFPA 805 were evaluated using a Fire Risk Evaluation per Section 4.2.4.2 of NFPA 805. A Fire Safety Analysis was performed for each fire area containing variances from the deterministic requirements of Section 4.2.3 of NFPA 805 (VFDRs), a Fire Risk Evaluation was performed for each fire area containing VFDRs HBRSEP LAR Rev 0 Page 34

Duke Energy 4.0 Compliance with NFPA 805 Requirements Ifthe Fire Risk Evaluation meets the acceptance criteria, this is confirmation that a success path effectively remains free of fire damage and that the performance-based approach is acceptable per Section 4.2.4.2 of NFPA 805.

The Fire Risk Evaluation process consists of the following steps (Figure 4-7 depicts the Fire Risk Evaluation process used during transition. This is generally based on FAQ 07-0054 Revision 1:

Step 1 - Preparation for the Fire Risk Evaluation.

Definition of the Variances from the Deterministic Requirements. The definition of the VFDR includes a description of problem statement and the section of NFPA 805 that is not met, type of VFDR (e.g., separation issue or degraded fire protection system), and proposed evaluation per applicable NFPA 805 section.

Preparatory Evaluation - Fire Risk Evaluation Team Review. Using the information obtained during the development of the NEI 04-02 B-3 Table and the Fire PRA, a team review of the VFDR was performed. Depending on the scope and complexity of the VFDR, the team may include the Safe shutdown/NSCA Engineer, the Fire Protection Engineer, and the Fire PRA Engineer. The purpose and objective of this team review was to address the following; o Review of the Fire PRA modeling treatment of VFDR o Ensure discrepancies were captured and resolved Step 2 - Performed the Fire Risk Evaluation The Evaluator coordinated as necessary with the Safe shutdown/NSCA Engineer, Fire Protection Engineer and Fire PRA Engineer to assess the VFDR using the Fire Risk Evaluation process to perform the following:

o Change in Risk Calculation with consideration for additional risk of recovery actions and required fire protection systems and features due to fire risk.

o Fire area change in risk summary Step 3 - Reviewed the Acceptance Criteria The acceptance criteria for the Fire Risk Evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are ACDF and ALERF. The qualitative factors are defense-in-depth and safety margin.

o Risk Acceptance Criteria. The transition risk evaluation was measured quantitatively for acceptability using the ACDF and ALERF criteria from RG 1.174, as clarified in RG 1.205 Regulatory Position 2.2.4.

o Defense-in-Depth. A review of the impact of the change on defense-in-depth was performed, using the guidance NEI 04-02. NFPA 805 defines defense-in-depth as:

- Preventing fires from starting

- Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage HBRSEP LAR Rev 0 Page 36

Duke Energy 4.0 Compliance with NFPA 805 Requirements Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the defense-in-depth requirement was considered to be satisfied if the proposed change does not result in a substantial imbalance among these elements (or echelons).

The review of defense-in-depth was qualitative and addressed each of the elements with respect to the proposed change. Defense-in-depth was performed on a fire area basis.

Fire protection features and systems relied upon to ensure defense-in-depth were identified as a result of the assessment of defense-in-depth.

o Safety Margin Assessment. A review of the impact of the change on safety margin was performed. An acceptable set of guidelines for making that assessment is summarized below. Other equivalent acceptance guidelines may also be used.

- 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., UFSAR, 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.

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Duke Energy 4.0 Compliance with NFPA 806 Requirements Prepare for Fire Risk Evaluation Determine How to Model Discuss and Document in the VFDR in the Fire PRA li 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]

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Duke Energy 4.0 Compliance with NFPA 805 Requirements Results of Evaluation Process Disposition of VFDRs The HBRSEP existing post-fire SSA and the NFPA 805 transition project activities have identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. These variances were dispositioned using the fire risk evaluation process.

Each variance dispositioned using a Fire Risk Evaluation was assessed against the Fire Risk Evaluation acceptance criteria of ACDF and ALERF; and maintenance of defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in Attachment C.

Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c).

Risk Change Due to NFPA 805 Transition In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using Fire Risk Evaluations and the overall plant should be provided. Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.

RG 1.205 Section C.2.2.4.2 states in part "The total 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 increasesresulting 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 additional risk associatedwith previously approved recovery actions is greaterthan the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurredby any proposed alternativesto the deterministic criteriain NFPA 805, Chapter4 (otherthan the previously approved recovery actions), should be risk neutral or representa risk decrease."

The risk increases and decreases are provided in Attachment W.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states:

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

Monitoring shall ensure that the assumptions in the engineeringanalysis remain valid."

As part of the transition review, the adequacy of the inspection and testing program to address fire protection systems and equipment within plant inspection and the compensatory measures programs should be reviewed. In addition, the adequacy of the plant corrective action program in determining the causes of equipment and programmatic failures and minimizing their recurrence should also be reviewed as part of the transition to a risk-informed, performance-based licensing basis.

4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program This section describes the process that will be utilized to implement the post-transition NFPA 805 monitoring program. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805.

See item for implementation in Attachment S. The monitoring process is comprised of four phases.

• Phase 1- Scoping

• Phase 2- Screening Using Risk Criteria

• Phase 3- Risk Target Value Determination

" Phase 4- Monitoring Implementation Figure 4-8 provides detail on the Phase 1 and 2 processes.

The results of these phases will be documented in the NFPA 805 Monitoring Program scoping document developed during implementation.

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Duke Energy 4.0 Compliance with NFPA 806 Requirements Phase I - Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program:

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

o Fire protection systems and features

- Required by the Nuclear Safety Capability Assessment

- Modeled in the Fire PRA

- Required by Chapter 3 of NFPA 805 o Nuclear Safety Capability Assessment equipment4

- Nuclear safety equipment

- Fire PRA equipment

- NPO equipment o Structures, systems and components relied upon to meet radioactive release criteria o Fire Protection Programmatic Elements Phase 2 - Screening Using Risk Criteria The equipment from Phase 1 scoping will be screened to determine the appropriate level of NFPA 805 monitoring. As a minimum, the SSCs identified in Phase 1 will be part of an inspection and test program and system/program health reporting. If not in the current program, the SSCs will be added in order to assure that the criteria can be met reliably.

The following screening process will be used to determine those SSCs that may require additional monitoring beyond normal inspection and test program and system/program health reporting and will be documented in NFPA 805 Monitoring Program scoping document.

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

Risk significance is determined at the component, programmatic element, and/or functional level on an individual fire area basis. Compartments smaller than fire areas may be used provided the compartments are independent (i.e., share no fire protection SSCs). If compartments smaller than fire areas are used the basis will be documented in the calculation, RNP-F/PSA-0095, RNP Fire PSA NFPA 805 Transition.

4 For the purposes of the NFPA 805 Monitoring, "NSCA equipment" is intended to include Nuclear Safety Equipment, Fire PRA equipment, and NPO equipment.

H13RSEP LAR Rev 0 Page 40

Duke Energy 4.0 Compliance with NFPA 805 Requirements 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) > 1.0E-7 per year (OR)

Large Early Release Frequency (LERF) x (RAW) > 1.OE-8 per year CDF, LERF, and RAW(monitored parameter) are calculated for each fire area. The 'monitored parameter' will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration) and will be documented in the calculation, RNP-F/PSA-0095.

Fire protection systems and features that meet or exceed the criteria identified above are considered High Safety Significant (HSS) and will be included in the NFPA 805 Monitoring Program. The HSS fire protection systems and features not already monitored via an existing inspection and test program and/or in the existing system I program health reporting, as described in procedure EGR-NGGC-0010, will be added to the NFPA 805 Monitoring Program and documented in the NFPA 805 Monitoring Program scoping document.

2. Nuclear Safety Capability Assessment Equipment Required NSCA equipment, except the NPO scope, identified in Phase 1 will be screened for safety significance using the Fire PRA and the Maintenance Rule 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 Maintenance Rule. All NSCA equipment that are not HSS are considered LSS and need not be included in the monitoring program.

For non-power operational modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement.

Therefore, fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire prevention programs. Additional monitoring beyond inspection and test programs and system/program health reporting is not considered necessary.

3. SSCs Relied upon for Radioactive Release Criteria The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSCs relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (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 HBRSEP LAR Rev 0 Page 41

Duke Energy 4.0 Compliance with NFPA 805 Requirements 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:

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

Therefore, monitoring is conducted using the existing 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 Failure criteria is established by an expert panel based on the required fire protection and nuclear safety capability SSCs and programmatic elements assumed level of performance in the supporting analyses established in Phase 2. Action levels are established for the SSCs at the component level, program level, or functionally through the use of the pseudo system or 'performance monitoring group' concept. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (i.e., 3 operating cycles).

Since the HSS NSCA equipment have been identified using the Maintenance Rule guidelines, the associated equipment specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions.

When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions. Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels.

Documentation of the monitoring program failure criteria and action level targets will be contained in a documented evaluation. It is anticipated that the availability and reliability criterion for High Safety Significant Performance Monitoring Groups will use the guidance included in several industry documents tempered by site-specific operating experience, Fire PRA assumptions, and equipment types (and vendor data or valid design input when available). Industry documents such as the EPRI Fire Protection Equipment Surveillance Optimization and Maintenance Guide TR-1006756, Final Report July 2003, NFPA codes, and/or the NRC Fire Protection Significance Determination Process in addition to site specific operating experience data may be used. The monitoring program failure criteria and action level targets will be documented in the NFPA 805 Monitoring Program scoping document.

HBRSEP LAR Rev 0 Page 42

Duke Energy 4.0 Compliance with NFPA 806 Reauirements 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 procedure, CAP-NGGC-0200 will be initiated to identify the negative trend. A corrective action plan will then be developed to ensure the performance returns to the established level.

When applicable, a sensitivity study can be performed to determine the margin below the action level that still provides acceptable fire PRA results to help prioritize corrective actions if the action level is reached.

A periodic assessment will be performed (i.e., at a frequency of approximately every two to three operating cycles), taking into account, where practical, industry wide operating experience. Issues that will be addressed include:

o 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?

o 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?

o Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?

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Duke Energy 4.0 Compliance with NFPA 805 Requirements Duke Energy 4.0 Compliance with NFPA 805 Requirements Figure 4 NFPA 805 Monitoring - Scoping and Screening Since the HSS SSCs have been identified using the Maintenance Rule guidelines, the associated SSC specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions. 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). 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, as described in FAQ 10-0059.

Page~

HBRSEPLARRevO HBRSEP LAR Rev 0 Page 44

Duke Energy 4.0 Compliance with NFPA 805 Requirements 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, HBRSEP has documented analyses to support compliance with 10 CFR 50.48(c). The analyses are being performed in accordance with Duke Energy's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analyses.

Analyses, as defined by NFPA 805 Section 2.4, performed to demonstrate compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy. Note these analyses do not include items such as periodic tests, hot work permits, fire impairments, etc.

The Fire Protection Design Basis Document described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 have been created as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation. Figure 4-9 shows the Planned Post-Transition Documents.

The Fire Protection licensing basis documents under NFPA 805 consist of the following:

• The Transition Report/LAR

" The NFPA805 SE

" The Revised License Condition

" The revised (U)FSAR The Fire Protection Program Design Basis Document (DBD) will contain or reference sub-tier documents that also form part of the fire protection program. The DBD's as described in NFPA 805 section 2.7.1.2, are the Fire Safety Analysis (FSA) calculations provided for each plant fire area. Also included, is the NFPA 805 Code Compliance Calculation which will maintain certain supporting elements of the LAR such as Tables B-1, B-2 and E-1. These and other supporting calculations are developed under fleet procedure EGR-NGGC-0017, Preparation and Control of Design Analyses and Calculations, and are maintained as design documents / controlled documents as described in the procedure.

Rev 0 Page 45 HBIRSEP LAR HBRSEP LAR Rev 0 Page 45

Duke Energy 4.0 Compliance with NFPA 805 Requirements NFPA 805 DOCUMENTS NSCA Database NSEL Comp Cables PRA Equipment NoPwe and Data Equipment and Data NSCA CALCULATION Comp & Cable FA Assessment Method/Results Method/Results Revised License Condition MSO and OMA SSA Drawings Treatments SADaig NSCA SUPPORTING INFO Revised UFSAR T-H Calculations MFeasibIlItyion

[ B 2Table IF Z 3 Tabsle FIRE SAFETY ANALYSIS (DBD)

Coordination Plant DBDs that " On a Fire Area Basis Calculations MHIFI Support NSCA -Fire Area Description MHIF -FHA Database information

-Nuclear Safety Performance Criteria Compliance Summary (NEI 04-02 B-3 Table Results)

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

" On a Generic Basis

• B-I Table Results Non-Power Operations Calculations

- Radioactive Release (Training)

'Monitoring Program NFPA 805 FIRE RISK EVALUATIONS Fire Risk Evaluation Calculation(s)

FHA DATABASE DATA Ignition Sources FP Systems and

& Scenarios Features Data Inventory of B-1 Table Hazards Detailed Data FHA SUPPORT DOCUMENTATION FP Systems Code Compliance FP Drawings Evaluations Bold text Indicates new NFPA 805 documents FP System and I FeaSstuem Engineering adB Equivalency Evaluations Radioactive Fire Pmo-Plans Release Review Calculation Figure 4 NFPA 805 Planned Post-Transition Documents and Relationships Page 46 HBRSEP LAR LAR Rev Rev 0 0 Page 46

Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805 Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) is subject to Duke Energy configuration control processes that meet the requirements of Section 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes impacting the fire protection program are reviewed appropriately. The RI-PB post transition change process methodology is based upon the requirements of NFPA 805, and industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-2.

Table 4-2 Change Evaluation Guidance Summary Table Document Section(s) Topic NFPA 805 2.2(h), 2.2.9, 2.4.4, A.2.2(h), A.2.4.4, Change Evaluation D.5 NEI 04-02 5.3, Appendix B,Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (Appendix I)

RG 1.205 C.2.2.4, C.3.1, C.3.2, C.4.3 Risk Evaluation, Standard License Condition, Change Evaluation Process, Fire PRA The Plant Change Evaluation Process consists of the following 4 steps and is depicted in Figure 4-10:

" Defining the Change

" Performing the Preliminary Risk Screening.

• Performing the Risk Evaluation

• Evaluating the Acceptance Criteria Configuration control is and will be maintained going forward in accordance with existing procedures and processes which satisfy the NFPA 805 requirements. Procedure FIR-NGGC-0010, Fire Protection Program Impact Review, provides review of configuration, process, and procedure changes to ensure applicable requirements of NFPA 805 Fire Protection Program (Fundamental Elements, NSCA, NPO, Radioactive Release and FPRA) are maintained.

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.

HBRSEP LAR Rev 0 Page 47

Duke Energy 4.0 Compliance with NFPA 805 Requirements Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-03 process. This process will address most administrative changes (e.g.,

changes to the combustible control program, organizational changes, 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.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements Defining the Change (5.3.2)

License N vt hp3o Amendment prvosyarve Request Atraie

, Yes 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 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 HEIRSEP Rev 0 LAR Rev HBRSEP LAIR 0 Page 49

Duke Energy 4.0 Compliance with NFPA 806 Requirements The HBRSEP Fire Protection Program configuration is defined by the program documentation. The existing configuration control processes for modifications, calculations and analyses, and Fire Protection Program License Basis Reviews will be utilized to maintain configuration control of the Fire Protection program documents. The configuration control procedures which govern the various HBRSEP documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements (Implementation Item in Attachment S).

Several NFPA 805 document types, such as NSCA Supporting Information, Non-Power Mode NSCA Treatment, generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. The new procedures will be modeled after the existing processes for similar types of documents and databases. System level design basis documents will be revised to reflect the NFPA 805 role that the system components now play.

The process for capturing the impact of proposed changes to the plant on the Fire Protection Program will continue to be a multiple step review. The first step of the review is an initial screening for process users to determine if there is a potential to impact the Fire Protection program as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used. Reviews that identify potential Fire Protection program impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, Fire PRA) to ascertain the program impacts, if any. If Fire Protection program impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:

" Deterministic Approach: Comply with NFPA 805, Chapter 3 and 4.2.3 requirements

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

This process follows the requirements in NFPA 805 and the guidance outlined in RG 1.174, which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. The plant documents that ensure these requirements are met are:

CAP-NGGC-0200 - Condition Identification and Screening Process EGR-NGGC-0005 - Engineering Change ESGO101 N - Safe Shutdown Engineer (Post-NFPA 805 Transition)

ESGO102N - Fire Protection Plant Change Impact Review ESGO1 03N - Circuit Analysis (Post-NFPA 805 Transition)

HBRSEP LAR Rev 0 Page 50

Duke Energy 4.0 Compliance with NFPA 805 Requirements ESGO104N - Fire Protection Engineer (Post-NFPA 805 Transition)

ESGO105N - Basic Fire Modeling 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality Duke Energy will maintain the existing Fire Protection Quality Assurance program.

During the transition to 10 CFR 50.48(c), HBRSEP performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.

Any future NFPA 805 analyses will be conducted in accordance with the Quality Requirements described in NFPA 805, section 2.7.3 under the design controls in place and required by the Fire Protection portions of the NGGM-PM-0007, Quality Assurance Program Manual.

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 Duke Energy 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. Duke Energy specifically requires that the calculations and evaluations in support of the NFPA 805 LAR, exclusive of the Fire PRA, be performed within the scope of the QA program which requires independent review as defined by plant procedures. As recommended by NUREG/CR-6850, the sources of uncertainty in the Fire PRA were identified and specific parameters were analyzed for sensitivity in support of the NFPA 805 Fire Risk Evaluation process.

Specifically with regard to uncertainty, an uncertainty and sensitivity matrix was developed and included with RNP-F/PSA-0094. In addition, sensitivity to uncertainty associated with specific Fire PRA parameters was quantitatively addressed in RN P-F/PSA-0095.

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 Duke Energy continues to strive toward a more "realistic" estimate of HBRSEP LAR Rev 0 Page 51

Duke Energy 4.0 Compliance with NFPA 805 Requirements 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 The following discusses how the requirements of NFPA 805 Section 2.7.3 were met during the transition process. Post-transition, Duke Energy will perform work in accordance with NFPA 805 Section 2.7.3 requirements.

Reference plant procedures:

EGR-NGGC-0003 - Design Review Requirements EGR-NGGC-0005 - Engineering Change EGR-NGGC-0017 - Preparation and Control of Design Analyses and Calculations Review and approval of corporate or fleet-wide procedures applied to HBRSEP and other Duke Energy Progress sites are controlled under PRO-NGGC-0204, Procedure Review and Approval. Site specific impact and technical reviews are completed under this process to ensure each individual plant's requirements and configurations are incorporated and maintained.

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 Duke Energy procedures that require independent review.

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.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements Post-transition, for personnel performing fire modeling or Fire PRA development and evaluation, Duke Energy 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. The following Training Guides have been developed and implemented.

ESGO089N - Fire Probabilistic Safety Assessment Engineer (Quantification),

ESGO093N - Fire Probabilistic Safety Assessment Engineer (Initial Development), and ESGO094N - Fire Probabilistic Safety Assessment Engineer (Data Development), and ESGO105N - Basic Fire Modeling HBRSEP and NGG Fleet engineering personnel (design, programs and systems engineering) are provided training commensurate with the job responsibility through the INPO accredited Engineering Support Personnel (ESP) training program. This is provided in either ESP Continuing Training or Work Group Specific Continuing Training.

Specific, qualification for performance of the FIR-NGGC-0010, FireProtection Program Change Process, is documented using Training Guide (Qual. Card) ESGO102N, Fire ProtectionPlant Change Impact Review.

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.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8 Summary of Results 4.8.1 Results of the Fire Area Review A summary of the NFPA 805 compliance basis and the required fire protection systems and features is provided in Attachment C. The table provides the following information from the NEI 04-02 Table B-3:

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

Description:

Fire Area/Zone Description.

" NFPA 805 Regulatory Basis: Post-transition NFPA 805 Chapter 4 compliance basis

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

The documentation of required fire protection systems and features does not include the documentation of the fire area boundaries. Fire area boundaries are required and documentation of the fire area boundaries has been performed as part of reviews of engineering evaluations, licensing actions, or as part of the reviews of the NEI 04-02 Table B-1 process. The basis for the requirement of the fire protection system / feature is designated as follows:

o S - Separation Criteria: Systems/Features required for Chapter 4 Separation Criteria in Section 4.2.3 o E - EEEE/LA Criteria: Systems/Features required for acceptability of Existing Engineering Equivalency Evaluations / NRC approved Licensing Action (i.e.,

Exemptions/Deviations/Safety Evaluations) (Section 2.2.7) o R - Risk Criteria: Systems/Features required to meet the Risk Criteria for the Performance-Based Approach (Section 4.2.4) o D - Defense-in-depth Criteria: Systems/Features required to maintain adequate balance of Defense-in-Depth for a Performance-Based Approach (Section 4.2.4)

Attachment W contains the results of the Fire Risk Evaluations, additional risk of recovery actions, and the change in risk on a fire area basis.

4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase Planned modifications, studies, and evaluations to comply with NFPA 805 are described in Attachment S.

In Attachment S, three tables are listed. Table S-1 identifies completed Plant Modifications, Table S-2 identifies Plant Modifications required to be completed.

Table S-3 identifies training, programs, personnel equipment, and document changes and upgrades required to be completed.

The Fire PRA model will represent the as-built, as-operated and maintained plant following completion of the risk related modifications identified in Attachment S. In the event the PRA model requires revision following completion of the modifications and HBRSEP LAR Rev 0 Page 54

Duke E neravy 4.0 Compliance with NFPA 805 Requirements implementation items listed in Attachment S, the changes will be controlled through normal HBRSEP processes. These changes are not expected to be significant.

4.8.3 Supplemental Information -Other Licensee Specific Issues 4.8.3.1 Fire PRA Qualitative Review The following methods and modeling aspects are qualitatively characterized as having a minimal impact or a conservative impact:

4.8.3.1.1 RCP Shutdown Seals The HBRSEP FPRA applied credit for the installation of the Westinghouse Shutdown Seal (SDS) to the Reactor Coolant Pumps as described in WCAP-17100-P-A. This credit provides a 98% reduction in the risk impact of a loss of RCP seal cooling based on a 2% failure rate for the SDS. Based on recent inspection results at other plants where the SDS modification has been implemented, the reliability of the SDS has been called into question. The intent is still to install the SDS at HBRSEP or an equivalent design in order to assure the credit assumed in the fire PRA is maintained.

4.8.3.1.2 NSCA Power Supply Strategy HBRSEP is changing the strategy used in a number of fire areas in the plant that were previously approved under the Appendix R licensing basis. The current strategy is to perform a fire incused load shed and recovery to limit operational uncertainties relative to circuit failure and spurious actuations. This strategy has been compared to what is often called self-induced station blackout (SISBO) in the industry. The strategy going forward is to use a symptom based operator response approach in response to fire events. The Attachment G recovery actions have been informed with the new strategy. The current fire PRA conservatively modeled the plant using the current load shed strategy. As part of program implementation, the operations procedures will be finalized and the Fire PRA updated. Therefore, there is a separate implementation item relative to this in Attachment S.

4.8.3.1.3 Hot Short Clearing For a selected set of risk significant Air Operated Valves (AOVs) and Solenoid Operated Valves (SOVs) that are considered to be 'failsafe' (i.e., fail in the desired position when hot shorts clear/power is removed), a determination was made whether damage would occur within the first 15 minutes in the event of a spurious hot short holding the valve in the undesired position. If 15 minutes could be survived, a probability of 0.06 was applied that the spurious hot short would clear. The hot short probability is expected to be congruent with or more conservative than future industry guidance.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.1.4 Control Power Transformers Impact on Circuit Failure Probabilities Circuit failure probabilities were calculated without credit for control power transformers.

The impact of control power transformers on circuit failure probabilities as described in industry guidance is currently being refined; therefore not applying credit provides conservative CDF and LERF results.

4.8.3.2 Fire PRA Sensitivity Analyses The following methods and modeling aspects are quantitatively evaluated. Each sensitivity analysis modifies an applied fire PRA method and then recalculates total risk metrics from Attachment W in order to determine the impact of a given fire PRA method.

The total plant risk and the total change in risk associated with the transition to NFPA 805 are reported for each sensitivity with the percentage change from the numbers reported in Attachment W.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.1 Closed MCC Sensitivity Analysis There are variations in methods applied to "closed" MCCs. If a cabinet were always "closed" there would be no fire impact on external targets. However, there is always the potential for the cabinet to already be open or an arc fault to have enough energy to open the cabinet. For the HBRSEP FPRA, it was assumed that one out of ten MCC fires may result in an "open" cabinet configuration. This is not applied to the HRR as a severity factor, but as a split fraction on the likelihood that the cabinet remains "closed."

The guidance for characterizing closed cabinets at the HBRSEP plant was the same as that used for the Harris Nuclear Plant pilot effort. The use of split fractions as described above was deemed acceptable for use by the NRC at HNP.

A sensitivity analysis was performed on this method for the HBRSEP plant fire PRA.

The sensitivity analysis essentially removed the split fraction, effectively treating the closed MCCs as always open. The results of the sensitivity for the "closed" cabinet method are provided below.

Table 4 Closed MCCs Sensitivity Delta CDF and Delta LERF Results ACDF [lyr] ALERF [Iyr]

(Change from (Change from baseline) baseline) 3.4E-06 3.6E-07 VFDRs

(-3%) (-20%)

1.9E-06 5.4E-07 Recovery Actions (+58%) (+59%)

5.3E-06 9.OE-07 Total

(+15%) (+13%)

Differences in percentages are due to rounding.

Table 4 Closed MCCs Sensitivity Total CDF and LERF Results CDF [Iyr] LERF [/yr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 2.6E-05 5.2E-06 Fire

(+26%) (+34%)

3.9E-05 7.1E-06 Total

(+15%) (+25%)

Differences in percentages are due to rounding.

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Duke Enerciy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.2 FAQ 08-0048 Sensitivity Analysis In order to use the updated fire bin ignition frequencies provided in Supplement 1 to NUREG/CR-6850, a sensitivity analysis must be performed comparing the impact of those bins characterized by an alpha from the EPRI TR-1 016735 analysis that is less than or equal to 1. While the new point estimates for the bin ignition frequencies better represent the data, uncertainties are large and a sensitivity analysis using the old frequencies was required to assess the potential impact of using the new frequencies.

The sensitivity was simplified by replacing all EPRI TR-1 016735 ignition frequencies with values from NUREG/CR-6850. This simplification is assumed to be conservative, as the ignition frequencies from NUREG/CR-6850 tend to be greater than those from EPRI TR-1016735. The results of the sensitivity for the EPRI ignition frequencies are provided below.

Table 4 Ignition Frequency Sensitivity Delta CDF and Delta LERF Results ACDF [/yr] ALERF [lyr]

(Change from (Change from baseline) baseline) 7.2E-06 8.9E-07 VFDRs

(+106%) (+98%)

1.7E-06 5.1E-07 Recovery Actions

_______________ (+42%) (+50%)

8.9E-06 1,4E-06 Total

(+93%) (+75%)

Differences in percentages are due to rounding.

Table 4 Ignition Frequency Sensitivity Total CDF and LERF CDF [lyr] LERF [lyr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 3.2E-05 5.8E-06 Fire

(+54%) (+52%)

4.5E-05 7.7E-06 Total

(+32%) (+35%)

Differences in percentages are due to rounding.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.3 Buses El and E2 High Energy Arcing Fault (HEAF) Sensitivity Analysis Analysis was performed to demonstrate that the HEAF on the El and E2 buses would not damage external targets. This sensitivity provides a comparison of the impact of the El/E2 HEAF damaging external targets on the overall CDF and LERF. The results of the sensitivity El and E2 HEAF are provided below.

Table 4 El and E2 HEAF Sensitivity Delta CDF and Delta LERF Results ACDF [/yr] ALERF [Iyr]

(Change from (Change from baseline) baseline) 3.4E-06 3.7E-07 VFDRs

(-3%) (-18%)

1.8E-06 4.9E-07 Recovery Actions (+50%) (+44%)

5.2E-06 8.6E-07 Total

._(+13%) (+7%)

Differences in percentages are due to rounding.

Table 4 El and E2 HEAF Sensitivity Total CDF and LERF CDF [Iyr] LERF [/yr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 2.7E-05 5.1E-06 Fire

(+27%) (+34%)

• 4.OE-05 7.OE-06 Total

(+18%) (+23%)

Differences in percentages are due to rounding.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.4 Low Voltage Cabinet Incipient Detection Sensitivity Analysis Incipient Fire Detection Systems are installed in select cabinets in the main control room, Hagan room, and Rod Control room. The Fire PRA applies guidance from FAQ 08-0046 when evaluating risk for incipient detection, also known as Very Early Warning Fire Detection Systems (VEWFDS).

For the sensitivity, it is assumed that the incipient detectors will be replaced with in-cabinet detectors, thus removing the credit of incipient detection as guided by FAQ 08-0046. The results of replacing the VEWFDS with normal in cabinet detection credit are below.

Table 4 Incipient In-Cabinet Detection Sensitivity Delta CDF and Delta LERF Results ACDF [/yr] ALERF [lyr]

(Change from (Change from baseline) baseline) 3.6E-06 4.3E-07 VFDRs

(+3%) (-4%)

1.5E-06 4.4E-07 Recovery Actions

_________________ (+25%) (+29%)

5.1E-06 8.7E-07 Total

(+11%) (+9%)

Differences in percentages are due to rounding.

Table 4 Incipient In-Cabinet Detection Sensitivity Total CDF and LERF Results CDF [/yr] LERF [/yr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 2.6E-05 4.8E-06 Fire

(+24%) (+24%)

3.9E-05 6.7E-06 Total

(+15%) (+18%)

Differences in percentages are due to rounding.

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Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.5 Cable Spread Room Area Wide Incipient Detection Sensitivity Analysis As permitted by Appendix P of NUREG/CR-6850 Volume 2, the FPRA credits the use of air-aspirated incipient detection, also known as Very Early Warning Fire Detection Systems (VEWFDS) for area wide detection in the cable spread room. The fire PRA credits prompt detection and suppression for the use of area-wide incipient detection.

Plant procedures will be developed and implemented to ensure that VEWFDS alarms are promptly addressed with qualified plant personnel who will be present in the immediate area prior to fire growth, allowing for fire prevention or prompt fire response.

The incipient detectors will be replaced with traditional spot detection for this sensitivity analysis. This removes the credit for the probabilities of prompt detection and prompt suppression as guided by Appendix P. The results of replacing the VEWFDS with spot detection credit are below.

Table 4 Area Wide Incipient Detection Sensitivity Delta CDF and Delta LERF Results ACDF [lyr] ALERF [Iyr]

(Change from (Change from baseline) baseline) 3.4E-06 3.6E-07 VFDRs

(-3%) (-20%)

1.7E-06 4.4E-07 Recovery Actions

________________(+42%) (+29%)

5.1E-06 8.OE-07 Total

(+11%) (0%)

Differences in percentages are due to rounding.

Table 4 Area Wide Incipient Detection Sensitivity Total CDF and LERF Results CDF [lyr] LERF [lyr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 2.3E-05 4.4E-06 Fire

(+10%) (+16%)

3.6E-05 6.3E-06 Total

(+6%) (+11%)

Differences in percentages are due to rounding.

Page 61 HBRSEP LAR HBRSEP LAR Rev Rev 0 0 Page 61

Duke Energy 4.0 Compliance with NFPA 805 Requirements 4.8.3.2.6 Incipient Stage Prevention for In-Cabinet Detection Sensitivity Analysis The FPRA credits the use of air-aspirated incipient detection, also known as Very Early Warning Fire Detection Systems (VEWFDS), in the Main Control Boards (MCBs),

Hagan Room cabinets, and Rod Control room for select cabinets. As per FAQ 08-0046, delta (6) (probability factor for failure of incipient-stage prevention credit) may be significantly reduced from a 100% chance of failure, given that VEWFDS, procedural changes and other preventative measures are implemented that improve locating in-cabinet fires. A 10% failure probability was assumed for incipient stage prevention, as HBRSEP will develop and implement incipient alarm procedures to facilitate appropriate plant response.

Table 4 Incipient Stage Prevention Sensitivity Delta CDF and Delta LERF Results ACDF [lyr] ALERF [lyr]

(Change from (Change from baseline) baseline) 3.7E-06 4.OE-07 VFDRs

(+6%) (-11%)

1.5E-06 4.1E-07 Recovery Actions

________________ (+25%) (+21%)

5.2E-06 8.1E-07 Total

(+13%) (+1%)

Differences in percentages are due to rounding.

Table 4 Incipient Stage Prevention Sensitivity Total CDF and LERF Results CDF [Iyr] LERF [/yr]

(Change from (Change from baseline) baseline)

Internal Events plus External Flooding and 1.3E-05 1.9E-06 High Winds 2.4E-05 4.1E-06 Fire

(+14%) (+8%)

3.7E-05 6.OE-06 Total

(+9%) (+5%)

Differences in percentages are due to rounding.

Page 62 HBRSEP LAR LAR Rev Rev 0 0 Page 62

Duke Energy 6.0 Regulatory Evaluation

5.0 REGULATORY EVALUATION

5.1 Introduction - 10 CFR 50.48 On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements.

10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition (NFPA 805), as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.

The voluntary adoption of 10 CFR 50.48(c) by HBRSEP does not eliminate the need to comply with 10 CFR 50.48(a) including the provision for nuclear plants licensed prior to 10 CFR 50, Appendix A, GDC 3, Fire Protection, becoming effective as is the case for HBRSEP 5. 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(o. Those regulatory requirementscontinue to apply to licensees that adopt NFPA 805. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirementsmay be met is different than under 10 CFR 50.48(b). Specifically, whereas GDC 3 refers to SSCs importantto safety, NFPA 805 identifies fire protection systems and features required to meet the Chapter I performance criteria through the methodology in Chapter4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii)requirement to limit fire damage to SSCs importantto safety so that the capabilityto safely shut down the plant is ensured is satisfied by meeting the performance criteria in Section 1.5.1 of NFPA 805. The Section 1.5.1 criteria include provisions for ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries,and process monitoring are achieved and maintained.

This methodology specifies a process to identify the fire protection systems and features 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 and qualification must meet any applicable requirements of NFPA 805, Chapter3. Having identified the requiredfire protection systems and features, the licensee selects either a deterministicor performance-basedapproach to 5 The General Design Criteria (GDC) in existence at the time HBRSEP was licensed (July, 1970) for operation were contained in Proposed Appendix A to 10CFR50, General Design Criteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967. (Appendix A to 10CFR50, effective in 1971 and subsequently amended, is somewhat different from the proposed 1967 criteria.) HBRSEP was evaluated with respect to the proposed 1967 GDC and the original FSAR contained a discussion of the criteria as well as a summary of the criteria by groups.

HBRSEP LAR Rev 0 Page 63

Duke Energy 6.0 Regulatory Evaluation demonstrate that the performance criteria are satisfied. This process satisfies the GDC 3 requirement to design and locate SSCs important to safety to minimize the probabilityand effects of fires and explosions." (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086)

The new rule provides actions that may be taken to establish compliance with 10 CFR 50.48(a), which requires each operating nuclear power plant to have a fire protection program plan that satisfies GDC 3, as well as specific requirements in that section. The transition process described in 10 CFR 50.48(c)(3)(ii) provides, in pertinent parts, that a licensee intending to adopt the new rule must, among other things, "modify the fire protection plan required by paragraph (a) of that section to reflect the licensee's decision to comply with NFPA 805." Therefore, to the extent that the contents of the existing fire protection program plan required by 10 CFR 50.48(a) are inconsistent with NFPA 805, the fire protection program plan must be modified to achieve compliance with the requirements in NFPA 805. All other requirements of 10 CFR 50.48 (a) and GDC 3 have corresponding requirements in NFPA 805.

A comparison of the current requirements in Appendix R with the comparable requirements in Section 3 of NFPA 805 shows that the two sets of requirements are consistent in many respects. This was further clarified in FAQ 07-0032, 10 CFR 50.48(a) and GDC 3 clarification (ML081400292). The following tables provide a cross reference of fire protection regulations associated with the post-transition HBRSEP fire protection program and applicable industry and HBRSEP documents that address the topic.

10 CFR 50.48(a)

Table 5-1 10 CFR 50.48(a) - ApplicabilitylCompliance Reference 10 CFR 50.48(a) Section(s) ApplicabilitylCompliance Reference (1) Each holder of an operating license issued under this See below part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part. This fire protection plan must:

(i) Describe the overall fire protection program for the NFPA 805 Section 3.2 facility; NEI 04-02 Table B-1 (ii) Identify the various positions within the licensee's NFPA 805 Section 3.2.2 organization that are responsible for the program; NEI 04-02 Table B-1 (iii) State the authorities that are delegated to each of NFPA 805 Section 3.2.2 these positions to implement those responsibilities; and NEI 04-02 Table B-1 (iv) Outline the plans for fire protection, fire detection NFPA 805 Section 2.7 and Chapters 3 and 4 and suppression capability, and limitation of fire NEI 04-02 B-1 and B-3 Tables damage.

(2) The plan must also describe specific features See below necessary to implement the program described in paragraph (a)(1) of this section such as:

(i) Administrative controls and personnel requirements NFPA 805 Sections 3.3.1 and 3.4 for fire prevention and manual fire suppression NEI 04-02 Table B-1 activities; Page 64 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page 64

Duke Enermy 5.0 Regulatory Evaluation Duke Ener4y 5.0iceiliytCmpiEvaluation Table 5-1 10 CFR 50.48(a) - Applicability/Compliance Reference 10 CFR 50.48(a) Section(s) ApplicabilitylCompliance Reference (ii) Automatic and manually operated fire detection and NFPA 805 Sections 3.5 through 3.10 and suppression systems; and Chapter 4 NEI 04-02 B-1 and B-3 Tables (iii) The means to limit fire damage to structures, NFPA 805 Section 3.3 and Chapter 4 systems, or components important to safety so that the NEI 04-02 B-3 Table capability to shut down the plant safely is ensured.

(3) The licensee shall retain the fire protection plan and NFPA 805 Section 2.7.1.1 requires that each change to the plan as a record until the documentation (Analyses, as defined by NFPA 805 Commission terminates the reactor license. The Section 2.4, performed to demonstrate compliance licensee shall retain each superseded revision of the with this standard) be maintained for the life of the procedures for 3 years from the date it was plant.

superseded. RDC-NGGC-0001 (4) Each applicant for a design approval, design Not applicable. HBRSEP is licensed under certification, or manufacturing license under part 52 of 10 CFR 50.

this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of appendix A to this part.

General Design Criterion 3 Table 5-2 GDC 3 - ApplicabilitylCompliance Reference GDC 3, Fire Protection, Statement ApplicabilitylCompliance 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 65 HBRSEP LAIR HBRSEP Rev 00 LAR Rev Page 65

Duke Energy 5.0 Regulatory Evaluation 10 CFR 50.48(c)

Table 5-3 10 CFR 50.48(c) - ApplicabilitylCompliance Reference 10 CFR 50.48(c) Section(s) ApplicabilitylCompliance Reference (1) Approval of incorporationby reference. National Fire Protection Association General Information.

(NFPA) Standard 805, "Performance-Based Standard for Fire Protection for NFPA 805 (2001 edition) is Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), the edition used.

which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51.

(2) Exceptions, modifications, and supplementation of NFPA 805. As used in General Information.

this section, references to NFPA 805 are to the 2001 Edition, with the NFPA 805 (2001 edition) is following exceptions, modifications, and supplementation: the edition used.

(i) Life Safety Goal, Objectives, and Criteria. The Life Safety Goal, The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed. Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.

(ii) Plant Damage/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. as documented in NEI 04-02 Table B-1.

(vi) Water supply and distribution. The italicized exception to Section 3.6.4 is HBRSEP complies as not endorsed. Licensees who wish to use the exception to Section 3.6.4 documented in Attachment A.

must submit a request for a license amendment in accordance with See NEI 04-02 Table B-i.

paragraph (c)(2)(vii) of this section.

HBRSEP LAR Rev 0 Page 66

Duke Energy 5.0 Regulatory Evaluation Table 5-3 10 CFR 50.48(c) - ApplicabilitylCompliance Reference 10 CFR 50.48(c) Section(s) Applicability/Compliance Reference (vii) Performance-based methods. Notwithstanding the prohibition in The use of performance-Section 3.1 against the use of performance-based methods, the fire based methods for NFPA 805 protection program elements and minimum design requirements of Chapter 3 is requested. See Chapter 3 may be subject to the performance-based methods permitted Attachment L.

elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

(3) Compliance with NFPA 805. See below (i) A licensee may maintain a fire protection program that complies with The LAR was submitted in NFPA 805 as an alternative to complying with paragraph (b) of this section accordance with for plants licensed to operate before January 1, 1979, or the fire protection 10 CFR 50.90. The LAR license conditions for plants licensed to operate after January 1, 1979. The included applicable license licensee shall submit a request to comply with NFPA 805 in the form of an conditions, orders, technical application for license amendment under § 50.90. The application must specifications/bases that identify any orders and license conditions that must be revised or needed to be revised and/or superseded, and contain any necessary revisions to the plant's technical superseded.

specifications and the bases thereof. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate. Any approval by the Director or the designee must be in the form of a license amendment approving the use of NFPA 805 together with any necessary revisions to the technical specifications.

(ii) The licensee shall complete its implementation of the methodology in The LAR and transition report Chapter 2 of NFPA 805 (including all required evaluations and analyses) summarize the evaluations and, upon completion, modify the fire protection plan required by paragraph and analyses performed in (a) of this section to reflect the licensee's decision to comply with NFPA 805, accordance with Chapter 2 of before changing its fire protection program or nuclear power plant as NFPA 805.

permitted by NFPA 805.

(4) Risk-informed or performance-based alternatives to compliance with No risk-informed or NFPA 805. A licensee may submit a request to use risk-informed or performance-based performance-based alternatives to compliance with NFPA 805. The request alternatives to compliance must be in the form of an application for license amendment under § 50.90 of with NFPA 805 (per this chapter. The Director of the Office of Nuclear Reactor Regulation, or 10 CFR 50.48(c)(4)) were designee of the Director, may approve the application if the Director or utilized. See Attachment P.

designee determines that the proposed alternatives:

(i) Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Page 67 HBRSEP LAR HBRSEP LAR RevRev 00 Page 67

Duke Energy 5.0 Regulatory Evaluation 5.2 Regulatory Topics 5.2.1 License Condition Changes The current HBRSEP fire protection license condition 3.E 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 HBRSEP conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded. HBRSEP determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the HBRSEP adoption of the new fire protection licensing basis.

5.2.3 Orders and Exemptions A review was conducted of the HBRSEP 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. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

• Involve a significant increase in the probability or consequences of an accident previously evaluated; or

• Create the possibility of a new or different kind of accident from any accident previously evaluated; or

" Involve a significant reduction in a margin of safety.

This evaluation is contained in Attachment Q.

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. HBRSEP has evaluated the proposed amendment and determined that it involves no significant hazards consideration.

Page 68 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page 68

Duke Energy 6.0 Regulatory Evaluation 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 UFSAR After the approval of the LAR, in accordance with 10 CFR 50.71 (e), the HBRSEP UFSAR will be revised. The format and content will be consistent with NEI 04-02 FAQ 12-0062. This will occur during the implementation phase. Changes to the UFSAR are controlled under procedure REG-NGGC-01 01, Final Safety Analysis Report Revisions.

5.5 Transition Implementation Schedule The following schedule for transitioning HBRSEP 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.

Modifications will be completed by the startup of the second refueling outage after issuance of the Safety Evaluation (SE). Appropriate compensatory measures will be maintained until modifications are complete.

Page 69 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page 69

Duke Energy 6.0 References

6.0 REFERENCES

The following references were used in the development of the TR. Additional references are in the NEI 04-02 Tables in the various Attachments.

Industry References

1. NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 2 Detailed Methodology," EPRI 1008239 Final Report, NUREG/CR-6850 / EPRI 1023259, Nuclear Regulatory Commission, Rockville, MD, September, 2005.

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

3. NUREG-1824, Volume 1, "V&V of Selected Fire Models for Nuclear Power Plant Applications Volume 1: Main Report," NUREG-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.

4. NUREG-1824, Volume 3, "Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 3: Fire Dynamics Tools (FDTS)," NUREG-1824/ EPRI 1011999, Salley, M. H. and Kassawara, R. P.,

NUREG-1824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., May, 2007.

5. NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide,"

Salley, M. H. and Kassawara, R. P., NUREG-1934/EPRI-1019195, U.S.

Nuclear Regulatory Commission, Office of Nuclear Reactor Research, Washington, D. C., November, 2012.

6. NUREG-1805, "Fire Dynamics Tools (FDTS)," lqbal, N. and Salley, M. H.,

NUREG-1805, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., October, 2004.

7. NEI 00-01, Guidance for Post-Fire Safe Shutdown Analysis

8. NEI 02-03, Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program

9. NEI 04-02, Guidance for Implementation of a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c), Rev. 2, 09-2005.

[ML0608800500]

10. NEI 04-06, Guidance for Self-Assessment of Circuit Failures

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

12. NIST SP 1026, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Technical Reference Guide," Jones, W. W., Peacock, HBRSEP LAR Rev 0 Page 70

Duke Energy 6.0 References R. D., Forney, G. P., and Reneke, P. A., National Institute of Standards and Technology, Gaithersburg, MD, April, 2009.

13. NIST SP 1041, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) User's Guide," Peacock, R. D., Jones, W. W., Reneke, P. A., and Forney, G. P., National Institute of Standards and Technology, Gaithersburg, MD, December, 2008.

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

Peacock, R. D., McGrattan, K., Klein, B., Jones, W. W., and Reneke, P. A.,

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

15. NRL/MR/6180-04-8746, "Verification and Validation Final Report for Fire and Smoke Spread Modeling and Simulation Support of T-AKE Test and Evaluation," Tatem, P.A., Budnick, E.K., Hunt, S.P., Trelles, J., Scheffey, J.L.,

White, D.A., Bailey, J., Hoover, J., and Williams, F.W., Naval Research Laboratory, Washington, DC, 2004.

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

17. 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.

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

19. Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current," Report Number 34, Building Research Institute, Tokyo, Japan, 1960.

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

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

22. SFPE Handbook of Fire Protection Engineering, Section 3-1, "Heat Release Rates," Babrauskas, V., The SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.

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

24. ASME/ANS Ra-Sa-2009, Addenda to ASME/ANS Ra-Sa-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers/American Nuclear Society, New York, HBRSEP LAR Rev 0 Page 71

Duke Energy 6.0 References

25. 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-1 824 / EPRI 1011999, Salley, M. H. and Kassawara, R. P., NUREG-1 824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., May, 2007.

26. EPRI Technical Report TR-1006756, EPRI Fire Protection Equipment Surveillance Optimization and Maintenance Guide Plant Specific References

1. AOP-014, Loss of CCW

2. AOP-022, Loss of Service Water

3. AOP-041, Response to Fire Event

4. CAP-NGGC-0200, Condition Identification and Screening Process

5. DSP-001, Alternate Shutdown Diagnostic

6. DSP-002, Hot Shutdown Using the Dedicated/Alternate Shutdown System

7. EDMG-001, Extreme Damage Event Early Actions

8. EDMG-002, Refueling Water Storage Tank (RWST)

9. EDMG-003, Condensate Storage Tank (CST)

10. EDMG-005, Containment Vessel (CV)

11. EDMG-01 1, Spent Fuel Pit Casualty

12. EDMG-012, Core Cooling Using Alternate Water Source

13. EDMG-013, Airborne Release Scrubbing

14. EGR-NGGC-0003, Design Review Requirements

15. EGR-NGGC-0005, Engineering Change

16. EGR-NGGC-0010, System & Component Trending Program and System Notebooks

17. EGR-NGGC-001 7, Preparation and Control of Design Analyses and Calculations

18. EPP-001, Loss of All AC Power

19. FIR-NGGC-0009, NFPA 805 Transient Combustibles and Ignition Source Controls Program

20. FIR-NGGC-0010, Fire Protection Program Change Process

21. FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment (NSCA)

HBRSEP LAR Rev 0 Page 72

Duke Energy 6.0 References

22. FPIP-0121, Radiological Release Reviews During Fire Fighting Activities

23. HBR2-0B060, Electrical Installation Practices, Notes and Details

24. RNP-M/MECH-1826, Hot Gas Layer Calculation

25. NED-M/MECH-1006, Generic Fire Modeling Treatments

26. NED-M/MECH-1007, Radiant Energy Target Damage Profile

27. NED-M/MECH-1008, Fire Zone of Influence Calculation

28. NED-M/MECH-1 009, Thermal Damage Time of Cables Above a Burning Ignition Source

29. OMA-NGGC-0203, Shutdown Risk Management

30. OMM-002, Fire Protection Manual

31. OMM-003, Fire Protection Pre-Plan

32. OMM-033, Implementation of CV Closure

33. OMP-003, Shutdown Safety Function Guidelines

34. PRO-NGGC-0204, Procedure Review and Approval

35. RDC-NGGC-0001, NGG Standard Records Management Program

36. REG-NGGC-0101, Final Safety Analysis Report Revisions

37. Report Number 0004-0042-412-002, Evaluation of Main Control Room Abandonment Times at the H.B. Robinson Nuclear Plant

38. Report Number 0004-0042-000-001, Evaluation of the Development and Timing of Hot Gas Layer Conditions in RNP Fire Zone 20

39. Report Number P2217-1021-01-01, Robinson Fire PRA Quantification Calculation

40. Report Number P2217-1021-01-03

41. RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Nuclear Plant

42. RNP-E/ELEC-1217, Non-Power Operations

43. RNP-F/PSA-0006, RNP Initiating Events Assessment

44. RNP-F/PSA-0014, Post Initiator Human Reliability Analysis

45. RNP-F/PSA-0018, PSA Model Appendix A - System Notebooks

46. RNP-F/PSA-0043, RNP PRA - Accident Sequence Notebook

47. RNP-F/PSA-0066, RNP Fire PRA Component Selection

48. RNP-F/PSA-0067, RPN Fire PRA Plant Partitioning and Ignition Frequency

49. RNP-F/PSA-0074, RNP Uncertainty Analysis

50. RNP-F/PSA-0077, RNP Quantification Calculation HBRSEP LAR Rev 0 Page 73

Duke Energy 6.0 References

51. RNP-F/PSA-0094, RNP Fire PSA Quantification

52. RNP-F/PSA-0095, RNP Fire PRA - NFPA 805 Transition Support

53. RNP-M/MECH-1826, Hot Gas Layer Calculation

54. RNP-M/MECH-1884, Verfication and Validation of Fire Models Supporting the Robinson Nuclear Plant (RNP) Fire PRA

55. SAM-i, Inject into the Steam Generator

56. SAM-3, Inject into the RCS

57. SAM-4, Inject into Containment

58. SAM-6, Control Containment Conditions

59. SAM-8, Flood Containment Page 74 HORSEP HBRSEP LAR Rev 0 LAR Rev 0 Page 74

Due neg Attachments Duke Enerav Attachments ATTACHMENTS Page 75 HBRSEP LAR Rev 00 LAR Rev Page 75

Duke Energy Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements 68 Pages Attached Page A-I HBRSEP LAR Rev 0 LAR Rev 0 Page A-1

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.1 General Chapter 3 Requirement: 3.1* General.

This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection systems and features. These fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard. Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein.

Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2 Fire Protection Plan Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.1 Intent Chapter 3 Reauirement: 3.2.1 Intent.

A site-wide fire protection plan shall be established. This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the plan's implementation. This section establishes the criteria for an integrated combination of components, procedures, and personnel to implement all fire protection program activities.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2 Management Policy Direction and Responsibility.

Chapter 3 Requirement: 3.2.2* Management Policy Direction and Responsibility.

HBRSEP LAR Rev 0 Page A-2

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements A policy document shall be prepared that defines management authority and responsibilities and establishes the general policy for the site fire protection program.

Compllance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.1 [Management Policy on Senior Management]

Chapter 3 Requirement: 3.2.2.1*

The policy document shall designate the senior management position with immediate authority and responsibility for the fire protection program.

Compliance Statement Comoliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3.1 &3.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.2 [Management Policy on Daily Administration]

Chanter 3 Reouirement: 3.2.2.2*

The policy document shall designate a position responsible for the daily administration and coordination of the fire protection program and its implementation.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3.7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.3 [Management Policy on Interfaces]

Chanter 3 ReQuirement: 3.2.2.3*

The policy document shall define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In addition, this policy document shall identity the various plant positions having the authority for implementing the various areas of the fire protection program.

HBRSEP LAR Rev 0 Page A-3

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.2.4 [Management Policy on AHJ]

Chapter 3 Requirement: 3.2.2.4*

The policy document shall identify the appropriate AHJ for the various areas of the fire protection program.

Comaliance Statement Compliance Basis Complies No Additional Clarification Rgeference Document Doc Details OMM-002,Fire Protection Manual Section 4.1.7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.2.3 Procedures Chapter 3 Requirement: 3.2.3* Procedures.

Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:

(1)

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

ComDliance Statement Comgliance Basis Main Header: Complies Main Header: No Additional Clarification Section (1):License Amendment Required Section (1): See Attachment L for Surveillance Optimization.

Reference Document Doc Details OMM-002,Fire Protection Manual ALL FP-012,Fire Protection Systems Minimum Equipment and Compensatory ALL Actions FP-013,Fire Protection Systems Surveillance Requirements ALL Chaoter 3 Requirement: 2)

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

Compllance Statement Comoliance Basis HBRSEP LAR Rev 0 Page A-4

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Section (2): Complies Section (2): No Additional Clarification Reference Document Doc Details FP-012,Fire Protection Systems Minimum Equipment and Compensatory ALL Actions OMM-002,Fire Protection Manual Section 8.13.2 Chaoter 3 Requirement: (3)

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

Compliance Statement Compliance Basis Section (3): Complies Section (3): No Additional Clarification Reference Document Doc Details EGR-NGGC-0008,Engineerng Programs ALL EGR-NGGC-0010,System & Component Trending Program and System Section 1.1 & Enclosure 1 Notebooks Chapter 3 Requirement: (4) Reviews of physical plant modifications and procedure changes for impact on the fire protection program.

Compliance Statement Compliance Basis Section (4): Complies Section (4): No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3 EGR-NGGC-0003,Design Review Requirements ALL EGR-NGGC-0005, Engineering Change ALL PRO-NGGC-0204,Procedure Review and Approval ALL EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL REG-NGGC-0010,10 CFR 50.59 AND SELECTED REGULATORY ALL REVIEWS Chaoter 3 ReQuirement: (5) Long-term maintenance and configuration of the fire protection program.

Compliance Statement Compliance Basis Section (5): Complies Section (5): No Additional Clarification Reference Document Doc Details EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL EGR-NGGC-0003,Design Review Requirements ALL EGR-NGGC-0005,Engineering Change ALL Chanter 3 Reauirement: (6) Emergency response procedures for the plant industrial fire brigade.

Compliance Statement Compliance Basis Section (6): Complies Section (6): No Additional Clarification Reference Document Doc Details FP-001 ,Fire Emergency ALL HBRSEP LAR Rev 0 Page A-5

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements AOP-041,Response to Fire Event ALL Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.3 Prevention Chapter 3 Requirement: 3.3 Prevention.

A fire prevention program with the goal of preventing a fire from starting shall be established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following:

(1) Prevention of fires and fire spread by controls on operational activities.

Compliance Statement Comoliance Basis Main Header: Complies Main Header: No Additional Clarification Section (1): Complies Section (1): No Additional Clarification Reference Document Dec Details OMM-002,Fire Protection Manual Section 8.4 FIR-NGGC-0003,Hot Work Permit ALL Chapter 3 Requirement: (2) Design controls that restrict the use of combustible materials The design control requirements listed in the remainder of this section shall be provided as described.

Comoliance Statement Compliance Basis Section (2): Complies Section (2): No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 8.3 FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION ALL SOURCE CONTROLS PROGRAM EGR-NGGC-0005,Engineering Change Attachment 3, Section A3.0.7 &Attachment 2, Section A3.3.24 Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.3.1 Fire Prevention for Operational Activities.

Chapter 3 Requirement: 3.3.1 Fire Prevention for Operational Activities.

The fire prevention program activities shall consist of the necessary elements to address the control of ignition sources and the use of transient combustible materials during all aspects of plant operations.

The fire prevention program shall focus on the human and programmatic elements necessary to prevent fires from starting or, should a fire start, to keep the fire as small as possible.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doec Details HBRSEP LAR Rev 0 Page A-6

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION ALL SOURCE CONTROLS PROGRAM OMM-002,Fire Protection Manual Section 8.4, ALL FIR-NGGC-0003,Hot Work Permit ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.1 General Fire Prevention Activities.

Chapter 3 Requirement: 3.3.1.1 General Fire Prevention Activities.

The fire prevention activities shall include but not be limited to the following program elements:

(1) Training on fire safety information for all employees and contractors including, as a minimum, familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms.

Compliance Statement Compliance Basis Main Header: Complies Main Header: No Additional Clarification Section (1): Complies Section (1): No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Attachment 10.1 & Section 3.16 GNI0008N,Initial General Employee Training - Contractors Computer Based Training (CBT)

GETSSG,General Employee Training Self Study Guide ALL GNB01 N,nitial General Employee Training - Progress Energy Personnel Computer Based Training (CBT)

Chaoter 3 Requirement: (2)

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

Comoliance Statement Comoliance Basis Section (2): Complies Section (2): No Additional Clarification Reference Document Doc Details FP-01 0,Housekeeping Controls Section 8.2 Attachment 10.2 Chanter 3 Requirement: (3)

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

Compliance Statement Compliance Basis Section (3): Complies Section (3): No Additional Clarification Reference Document Doc Details EGR-NGGC-0005,Engineering Change Attachment 3, Section A3.0.7 & Attachment 2, Section A3.3.24 OMM-002,Fire Protection Manual Section 3.8.2 WCP-NGGC-0300,Work Request Initiation, Screening, Prioritization, and Section 9.2.1 .h Classification EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL HBRSEP LAR Rev 0 Page A-7

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.2 Control of Combustible Materials Chaoter 3 Reauirement: 3.3.1.2" Control of Combustible Materials.

Procedures for the control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:

(1)

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

Exception: Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 cm) or larger shall not be required to be fire-retardant treated.

Compliance Statement Compliance Basis Main Header: Complies Main Header: No Additional Clarification Section (1): Complies Section (1): No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.1.8 SOURCE CONTROLS PROGRAM FP-010,Housekeeping Controls ALL Chapter 3 Reauirement: (2) Plastic sheeting materials used in the power block shall be fire-retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale tests, or equivalent.

Compliance Statement Compliance Basis Section (2): Complies Section (2): Complies Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.1.9 SOURCE CONTROLS PROGRAM Chapter 3 Reauirement: (3) Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area immediately following the completion of work or at the end of the shift, whichever comes first.

Compliance Statement Compliance Basis Section (3): Complies Section (3): No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.1.2 SOURCE CONTROLS PROGRAM Chapter 3 Requirement: (4)

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

Compliance Statement Compliance Basis Section (4): Complies Section (4): No Additional Clarification HBRSEP LAR Rev 0 Page A-8

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details FP-Ol OHousekeeping Controls Attachments 10.1-10.3 Chanter 3 Requirement: (5)

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

Comoliance Statement Compliance Basis Section (5): Complies Section (5): No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.4 SOURCE CONTROLS PROGRAM FAQ 06-0020,Identification of "applicable NFPA standards" ALL UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-9 Chapter 3 Requirement: (6)

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

Compliance Statement Compliance Basis Section (6): Complies Section (6): No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION ALL SOURCE CONTROLS PROGRAM FP-006,Handling of Flammable Liquids and Gases ALL FAQ 06-0020,Identification of "applicable NFPA standards" ALL Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.3.1.3 Control of Ignition Sources Chapter 3 Reauirement: 3.3.1.3 Control of Ignition Sources Compliance Statement Comoliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chaater 3

Reference:

3.3.1.3.1 [Control of Ignition Sources Code Requirements]

Chapter 3 Reauirement: 3.3.1.3.1*

A hot work safety procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.

Compliance Statement Compliance Basis Complies COMPLIES: No Additional Clarification HBRSEP LAR Rev 0 Page A-9

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Complies with Clarification COMPLIES WITH CLARIFICATION:

Compliance with NFPA 241 is by clarification and is addressed through compliance with NFPA 51B. NFPA 241, 2009 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot work, states "Responsibility for hot work operations and fire prevention precautions, including permits and fire watches, shall be in accordance with NFPA 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work."

Reference Document Doc Details FIR-NGGC-0003,Hot Work Permit ALL NFPA 241,Standard for Safeguarding Construction, Alteration, and Section 5.1 Demolition Operations, 2004 Edition NED-M/BMRK-0001 ,Code Compliance Evaluation for NFPA 51 B, Standard ALL for Fire Prevention during Welding, Cutting, and Other Hot Work- 1999 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.2 [Control of Ignition Sources on Smoking Limitations]

Chapter 3 Reouirement: 3.3.1.3.2 Smoking and other possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant.

Comoliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details NO-80-169,Revision to the Administrative Controls for Fire Protection, Enclosure No. 4, Page 2 2/1/1980 FP-010,Housekeeping Controls Section 5.3 FIR-NGGC-0003,Hot Work Permit ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.3 [Control of Ignition Sources for Leak Testing]

Chapter 3 Requirement: 3.3.1.3.3 Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing Compliance Statement Compliance Basis HBRSEP LAR Rev 0 Page A-1 0

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0003,Hot Work Permit Section 6.15 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.1.3.4 [Control of Ignition Sources on Portable Heaters]

Chanter 3 Requirement: 3.3.1.3.4*

Plant administrative procedure shall control the use of portable electrical heaters in the plant. Portable fuel-fired heaters shall not be permitted in plant areas containing equipment important to nuclear safety or where there is a potential for radiological releases resulting from a fire.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.1.11 SOURCE CONTROLS PROGRAM FIR-NGGC-0003,Hot Work Permit Section 6.13 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.2 Structural.

Chapter 3 Reauirement: 3.3.2 Structural.

Walls, floors, and components required to maintain structural integrity shall be of noncombustible construction, as defined in NFPA 220, Standard on Types of Building Construction.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details UFSARHBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1.B-6 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.3 Interior Finishes Chapter 3 Requirement: 3.3.3 Interior Finishes.

Interior wall or ceiling finish classification shall be in accordance with NFPA 101, Life Safety Code@,

requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class I interior floor finishes.

Compliance Statement Compliance Basis HBRSEP LAR Rev 0 Page A-1 1

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Complies No Additional Clarification Reference Document Doc Details NFPA 101,Life Safety Code, 2009 Edition (a) Sections 10.2.3.4 & 10.2.7.4 CPL-XXXX-W-005,Nuclear Power Plant Protective Coatings ALL L2-C-007,Field Coatings ALL UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)

FPP-RNP-900,Fire Hazards Analysis Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.3.4 Insulation Materials Chaoter 3 Reauirement: 3.3.4 Insulation Materials.

Thermal insulation materials, radiation shielding materials, ventilation duct materials, and soundproofing materials shall be noncombustible or limited combustible.

Comoliance Statement Comoliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1 .B-6 CPL-HBR2-M-025,Heating, Ventilation, and Air Conditioning (HVAC) Main Section 11-2.01 Plant Fabrication and Installation CPL-HBR2-M-028,Specification for RHR Pump Pit to HVE-5 Exhaust Tie-In Section 11-2.01 Fabrication and Installation L2-M-039,Piping and Equpment Thermal Insulation Section 4.4.1.4 GID/87038-0014,Fire Barrier System ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5 Electrical.

Chaoter 3 Reouirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Charter 3

Reference:

3.3.5.1 [Electrical Wiring Above Suspended Ceiling Limitations]

Chaoter 3 Reouirement: 3.3.5.1 HBRSEP LAR Rev 0 Page A-1 2

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Wirng above suspended ceiling shall be kept to a minimum. Where installed, electrical wiring shall be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers.

Comolliance Statement Comoliance Basis License Amendment Required NRC approval is being requested in Attachment L for electrical wiring above suspended ceilings that may not comply with the requirements of NFPA 805.

Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-7 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.5.2 [Electrical Raceway Construction Limits]

Chaoter 3 Reguirement: 3.3.5.2 Only metal tray and metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components.

Compliance Statement Compliance Basis License Amendment Required NRC approval is being requested in Attachment L for electrical raceway construction at HBRSEP that may not comply with the requirements of NFPA 805.

Reference Document Doc Details HBR2-0B060 Sht D6,Electrical Installation Practices, Notes and Details ALL HBR2-0B060 SH D2,Electrical Installation Practices, Notes and Details ALL HBR2-0B060 SHC3,ELECTRICAL INSTALLATION PRACTICES, NOTES ALL AND DETAILS DBD/R87038/SD62,Design Basis Document Cable and Raceway System ALL Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.3.5.3 [Electrical Cable Flame Propagation Limits]

Chapter 3 Requirement: 3.3.5.3*

Electric cable construction shall comply with a flame propagation test as acceptable to the AHJ.

Comoliance Statement Compliance Basis Complies with Clarification COMPLIES WITH CLARIFICATION: FAQ 06-0022 provides an Appendix to evaluate Complies via Previous NRC Approval currently recognized flame propagation tests to the IEEE 383-1974 Standard, the Complies via Engineering Evaluation US NRC minimum test standard, and HBRSEP LAR Rev 0 Page A-1 3

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements acceptance criteria for cable flame propagation tests.

COMPLIES VIA PREVIOUS NRC APPROVAL: In the SER dated 2/28/78, the NRC stated the following:

"4.8 Electrical Cables In the plant areas outside the containment, cable jacket and insulation material is polyvinyl chloride. Inside the containment, cable insulation is silicone rubber. The flame test standard for cables IEEE 383 was not in effect at the time electrical cables were purchased and installed at H. B. Robinson. Cables in critical areas, inside and outside containment will be coated with a flame retardant coating. Detailed discussion of these areas can be found in Section 5.0 of this report."

Section 5.0 listed the following areas where cables would be coated:

Safety Injection Pump Room (Fire Area No. 3)

Component Cooling Water Pump Room (Fire Area No. 5)

Aux Feedwater Pump room (Fire Area No.

7)

Cable Vaults (Fire Areas No. 9 and No. 34)

Aux Building Hallway, Lower Level (Fire Area 10A, 10B, 10C)

Aux Building Hallway, Upper Level (Fire Areas 14A thru 14G, except 14D)

Unit 2 Cable Spreading Room, Computer Room (Fire Area No. 18)

Electrical Equipment Area (Fire Area No.

19)

Rod Control Room (Fire Area No. 20)

In the evaluation dated 2/21/80, the NRC stated:

"3.1.4 Fire Retardant Cable Coating Fire retardant coating will be applied to cables located in 13 different fire areas of the plant (4.8)."

By letter dated December 5, 1978, the licensee stated that the flame-retardant coating would be applied in accordance with manufacturers recommendations and that the manufacturer would be consulted to determine alternate HBRSEP LAR Rev 0 Page A-14

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements application methods for situations not covered by the manufacturer's standard recommendations.

We accept the licensee's proposal."

Per the 2/21/80 evaluation, the status of Fire Retardant Coating was "Complete".

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the basis for approval.

COMPLIES VIA ENGINEERING EVALUATION: Engineering evaluations EE-84-0043, EE-90-0037, EE-92-0090, and NED-B/BOP-1001 are applicable to electric cable construction at HBRSEP.

Reference Document Doc Details NLU-78-71 ,License Amendment 31 Section 4.8 & 5.0 NLU-80-106,RFI and Requirements to Resolve Issues Concerning Fire Section 3.1.4 Protection UFSARHBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.4.4.2 FAQ 06-0022,Acceptable Electrical Cable Construction Tests ALL DBD/R87038/SD62,Design Basis Document Cable and Raceway System Sections 3.5.1.3.3 and 3.5.1.3.5 EE-84-0043,Qualification of Rockbestos Fire Zone R Cable to IEEE-383 ALL Vertical Flame Test EE-90-0037,Evaluation Of Use Of Non-IEEE 383, Vertical Flame Test ALL Cable Proposed By Modification M-1001 EE-92-0090,Evaluation Of Abandoned Cables (Belden # 8424) Inside ALL Containment (General Area)

NED-B/BOP-1001 ,Comparative Analysis of Fire Propagation Characteristics ALL of UL-910 and IEEE-383 Qualified Cables Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.6 Roofs.

Chaoter 3 Reauirement: 3.3.6 Roofs.

Metal roof deck construction shall be designed and installed so the roofing system will not sustain a self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building.

Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods of HBRSEP LAR Rev 0 Page A-1 5

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Fire Tests of Roof Coverings.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-6 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7 Bulk Flammable Gas Storage.

Chapter 3 Reauirement: 3.3.7 Bulk Flammable Gas Storage.

Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FP-O1 O,Housekeeping Controls Section 8.1.7 UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8 EGR-NGGC-0005,Engineering Change ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.1 [Bulk Flammable Gas Location Requirements]

Chapter 3 ReQuirement: 3.3.7.1 Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire or explosion will not adversely impact systems, equipment, or components important to nuclear safety.

NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed for hydrogen storage.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 50A as evaluated in RNP-M/BMRK-1015.

Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8 RNP-MIBMRK-1015,Code Compliance Evaluation for NFPA 50A, Standard ALL for Gaseous Hydrogen Systems at Consumer Sites - 1999 Edition Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.2 [Bulk Flammable Gas Container Restrictions]

HBRSEP LAR Rev 0 Page A-1 6

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chapter 3 Reauirement: 3.3.7.2 Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is not pointed at buildings.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8 SAF-SUBS-00023,Compressed Gases Section 5.i.8 MCP-NGGC-0402,Material Management (Storage, Issue, and Maintenance) Section 9.1.6 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.7.3 [Bulk Flammable Gas Cylinder Limitations]

Chaoter 3 Reauirement: 3.3.7.3 Flammable gas storage cylinders not required for normal operation shall be isolated from the system.

Comoliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details SAF-NGGC-2172,Industrial Safety Section 9.14 SAF-SUBS-00023,Compressed Gases Section 5.g.5 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.8 Bulk Storage of Flammable and Combustible Liquids.

Chapter 3 Reauirement: 3.3.8 Bulk Storage of Flammable and Combustible Liquids.

Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing systems, equipment, or components important to nuclear safety. As a minimum, storage and use shall comply with NFPA 30, Flammable and Combustible Liquids Code.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 30 as evaluated in RNP-M/BMRK-1002.

Reference Document Doc Details RNP-M/BMRK-1002,Code Compliance Evaluation NFPA 30 - Unit 2 Diesel ALL Fuel Oil Storage Tanks FP-010,Housekeeping Controls Section 8.1.7, Attachment 10.1 UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-9 FP-006,Handling of Flammable Liquids and Gases ALL HBRSEP LAR Rev 0 Page A-1 7

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.9 Transformers.

Chapter 3 Reauirement: 3.3.9* Transformers.

Where provided, transformer oil collection basins and drain paths shall be periodically inspected to ensure that they are free of debris and capable of performing their design function.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OST-642,Main Transformer Deluge System Flow Test (Refueling Interval) ALL Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.3.10 Hot Pipes and Surfaces.

Chapter 3 ReQuirement: 3.3.10* Hot Pipes and Surfaces.

Combustible liquids, including high flashpoint lubricating oils, shall be kept from coming in contact with hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation.

Compnlance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FP-010,Housekeeping Controls Section 8.1 FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.4.1.1 SOURCE CONTROLS PROGRAM Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.3.11 Electrical Equipment Chapter 3 Requirement: 3.3.11 Electrical Equipment Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment.

Comoliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND IGNITION Section 9.4.12 SOURCE CONTROLS PROGRAM HBRSEP LAR Rev 0 Page A-1 8

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3 Reference; 3.3.12 Reactor Coolant Pumps.

Chapter 3 Reauirement: 3.3.12* Reactor Coolant Pumps.

For facilities with non-inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.

(1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil system.

Compliance Statement ComDrlance Basis Complies via Previous NRC Approval No oil collection system is provided for the reactor coolant pumps at HBRSEP.

Section (1): Complies via Previous NRC Approval By letter NLS-85-176 (3/7/1985), in response to HBRSEP request for exemption for requiring reactor coolant pump oil collection systems, the NRC stated the following:

"The containment contains three reactor coolant pumps (A, B and C). These are located in bays (A, B and C). These bays also contain safety related cabling for the reactor coolant loop instrumentation. Bays A and B share a common ceiling; Bay C is isolated from Bays A & B to some extent.

The bays are covered by removable concrete blocks. These blocks will cause the plume from an unmitigated fire to be diverted through the steam generator area. This area contains safety related steam flow instrumentation sensing lines.

Oil spilled in Bay A, will be confined to Bay A; however, oil spilled in Bays B and C can flow to adjacent areas. The foundation for the reactor coolant pumps is at the 237.000' level. The foundation for the steam generators is at the 238.33' level. The reactor coolant pump is located between the pressurized portion of the oil system and the steam generator supports, and serves to shield the steam generator supports in the event of an oil system rupture.

The major combustible in each bay is the 200 gallons of oil in each reactor coolant HBRSEP LAR Rev 0 Page A-1 9

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements pump.

The existing fire detection system in each reactor coolant pump bay is a two-zone detection system. One zone consists of a single infrared flame detector; the other zone consists of a 325°F fixed-temperature heat detector. Activation of one zone of detection sends an alarm to the control room; activation of the second zone of detection alarms in the control room and also opens the preaction water deluge valve to the bay. Both detectors are wall mounted.

The existing fire suppression system for each bay, is a preaction sprinkler system.

Each bay has its own deluge valve, supply header, and a ring header that encircles the reactor coolant pumps at elevation 239 feet 4 inches.

Each of the five risers off the ring header have three 220°F closed head side wall sprinklers at approximately 240 feet, 245 feet and 252 feet. elevations. These systems are designed to meet the minimum residual pressure and flow requirements of NFPA-Std-15.

The suppression system ring header piping in Bay A is designed to withstand an SSE, while Bays B and C are designed such that a seismic event would not impact safety related equipment due to suppression system rupture. The risers are restrained to withstand the nozzle reaction forces. These forces are greater than those anticipated from a seismic event.

The existing containment spray system would be used as an emergency back-up to the bay suppression system if necessary to cool the operating level and containment annulus outside of the RCP bays.

By letter dated June 7, 1983, the licensee proposed to:

(1) Provide additional ceiling mounted heat detectors to meet the spacing and location requirements of NFPA-STD-HBRSEP LAR Rev 0 Page A-20

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements 72E, "Standard on Automatic Fire Detectors.

(2) Replace existing closed head sprinklers with special open water spray nozzles and manual actuation from the control room.

(3) Construct 6 inch dikes at the 231 feet elevation in Bay B and Bay C.

(4) Revise operating procedures for the containment spray system to allow its operation as a back up fire suppression system with the sodium hydroxide valves out.

By letter dated October 5, 1983, the licensee committed to maintain an automatically actuated closed-head preaction system in lieu of a manually actuated open-head system.

We have evaluated the fire protection for the reactor coolant pump lube oil system and conclude that the effects of a fire in an RCP Bay will not prevent safe shutdown capability. There are no components within the RCP Bay that are required for safe shutdown. The effects of any fire within an RCP Bay will be prevented from affecting the safe shutdown equipment outside the RCP Bay by the suppression system inside the RCP Bay and the Containment Spray System outside the Bay.

It is the staffs conclusion that: 1) installation of a reactor coolant pump oil collection system in this facility would not significantly enhance fire safety, and 2) the existing fire protection system in the Reactor Coolant Pump Bays with the addition of the proposed modifications provides an acceptable level of safety to that achieved by compliance with the requirements of Section 111.0 of Appendix R to 10 CFR 50. Therefore, the licensee's request for an exemption should be granted."

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

HBRSEP LAR Rev 0 Page A-21

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements There have been no plant modifications or other changes that would invalidate the basis for approval.

Section (1): No oil collection system is provided for the reactor coolant pumps at RNP. See Section 3.3.12 above for discussion of acceptability for lack of oil collection system.

Reference Document Doc Details NLS-85-176,RCP Oil Collection Exemption ALL Chapter 3 Requirement: (2) Leakage shall be collected and drained to a vented closed container that can hold the inventory of the reactor coolant pump lubricating oil system.

Compliance Statement Compliance Basis Section (2): Complies via Previous NRC Section (2): No oil collection system is Approval provided for the reactor coolant pumps at HBRSEP. See Section 3.3.12 above for discussion of acceptability for lack of oil collection system.

Chapter 3 Requirement: (3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazard of a fire flashback.

Compliance Statement Compliance Basis Section (3): Complies via Previous NRC Section (3): No oil collection system is Approval provided for the reactor coolant pumps at HBRSEP. See Section 3.3.12 above for discussion of acceptability for lack of oil collection system.

Chanter 3 Reauirement: (4) Leakage points on a reactor coolant pump motor to be protected shall include but not be limited to the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps.

Compliance Statement Compliance Basis Section (4): Complies via Previous NRC Section (4): No oil collection system is Approval provided for the reactor coolant pumps at HBRSEP. See Section 3.3.12 above for discussion of acceptability for lack of oil collection system.

Chapter 3 Reouirement: (5) The collection basin drain line to the collection tank shall be large enough to accommodate the largest potential oil leak such that oil leakage does not overflow the basin.

Comoliance Statement Compliance Basis Section (5): Complies via Previous NRC Section (5): No oil collection system is Approval provided for the reactor coolant pumps at HBRSEP. See Section 3.3.12 above for discussion of acceptability for lack of oil HBRSEP LAR Rev 0 Page A-22

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements collection system.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4 Industrial Fire Brigade.

Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.1 On-Site Fire-Fighting Capability.

Chapter 3 Requirement: 3.4.1 On-Site Fire-Fighting Capability.

All of the following requirements shall apply.

(a) A fully staffed, trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable:

(1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting)

Compliance Statement Compliance Basis Section (a): Complies Section (a): No Additional Clarification Section (a) (1): Complies via Engineering Section (a) (1): HBRSEP complies with Evaluation NFPA 600 as evaluated in NED-M/BMRK-0002.

Reference Document Doc Details OMM-002,Fire Protection Manual Section 8.6 NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR NFPA 600, ALL STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000 EDITION Chanter 3 Requirement: (2) NFPA 1500, Standard on Fire Department Occupational Safety and Health Program Compliance Statement Comollance Basis Section (a) (2): N/A Section (a) (2): NFPA 1500 is not applicable to HBRSEP as the site utilizes a fire brigade, not an organized fire department. Fire Brigade requirements are reviewed using NFPA 600.

Chapter 3 Requirement: (3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for Fire Department Physicians.

Compliance Statement Comoliance Basis Section (a) (3): N/A Section (a) (3): NFPA 1582 is not HBRSEP LAR Rev 0 Page A-23

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements applicable to HBRSEP as the site utilizes a fire brigade, not an organized fire department. Fire Brigade requirements are reviewed using NFPA 600.

Chaoter 3 Reauirement: (b)

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

Compliance Statement Comollance Basis Section (b): Complies Section (b): No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 3.12 & 8.6 Chapter 3 Requirement: (c) During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance Exception: Sufficient training and knowledge shall be permitted to be provided by an operations advisor dedicated to industrial fire brigade support criteria.

Compliance Statement Compliance Basis Section (c): Complies Section (c): No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 8.6.1 FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL Chaoter 3 Reouirement: (d)

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

Compliance Statement Compliance Basis Section (d): Complies Section (d): No Additional Clarification Reference Document Doc Details AOP-041,Response to Fire Event Section 2 Step 4 Chapter 3 Reauirement: (e) Each industrial fire brigade member shall pass an annual 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.

Compliance Statement Compliance Basis Section (e): Complies Section (e): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.4.1 OMM-002,Fire Protection Manual Section 8.6.3 Table B-1 NFPA 805 Ch.3 Transition Details HBRSEP LAR Rev 0 Page A-24

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chapter 3

Reference:

3.4.2 Pre-Fire Plans.

Chapter 3 Reauirement: 3.4.2* Pre-Fire Plans.

Current and detailed pre-fire plans shall be available to the industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Sections 9.2 and 9.5 OMM-002,Fire Protection Manual Section 8.8 HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60 ALL Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.4.2.1 [Pre-Fire Plan Contents]

Chapter 3 Reauirement: 3.4.2.1*

The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area, along with any nuclear safety components and fire protection systems and features that are present.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 9.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2.2 (Pre-Fire Plan Updates]

Chapter 3 Reauirement: 3.4.2.2 Pre-fire plans shall be reviewed and updated as necessary.

Comollance Statement Comnliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 4.0 AP-043,RNP Procedure Biennial Review Process Section 8.1 EGR-NGGC-0005,Engineering Change Section 4.5 RNP/94-1890,PROPOSED CHANGE TO QUALITY ASSURANCE ALL PROGRAM Table B-1 NFPA 805 Ch.3 Transition Details HBRSEP LAR Rev 0 Page A-25

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chanter 3

Reference:

3.4.2.3 [Pre-Fire Plan Locations]

Chapter 3 Requirement: 3.4.2.3*

Pre-fire plans shall be available in the control room and made available to the plant industrial fire brigade.

Comp)liance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 9.5 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.2.4 [Pre-Fire Plan Coordination Needs]

Chaoter 3 Reouirement: 3.4.2.4*

Pre-fire plans shall address coordination with other plant groups during fire emergencies.

Compliance Statement Compliance Basis Complies with Clarification HBRSEP has procedure FP-001, "Fire Emergency" which is not specifically a fire pre-plan, however FP-001 provides specific instructions for actions required from key groups at HBRSEP supporting the fire brigade/fire emergency actions.

There are detailed response coordination actions specified for Control Room, RC, and the Security group. Any other coordination actions would be initiated by the Control Room personnel as needed for any plant emergency.

Reference Document Doc Details FP-001 ,Fire Emergency ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.3 Training and Drills.

Chapter 3 Renuirement: 3.4.3 Training and Drills.

Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.

(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.

(1) Plant industrial fire brigade members shall receive training consistent with the requirements contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate.

Compliance Statement Compliance Basis HBRSEP LAR Rev 0 Page A-26

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Main Header: Complies Main Header: No Additional Clarification Section (a) (1): Complies via Engineering Section (a) (1): HBRSEP complies with Evaluation NFPA 600 as evaluated in the applicable portions of NED-M/BMRK-0002.

Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR NFPA 600, ALL STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000 EDITION Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(2) Industrial fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire.

Compliance Statement Compliance Basis Section (a) (2): Complies Section (a) (2): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL GNR01 N,Plant Access Annual Requalification, CBT ALL Chapter 3 Reouirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(3) A written program shall detail the industrial fire brigade training program.

Compliance Statement Compliance Basis Section (a) (3): Complies Section (a) (3): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(4) Written records that include but are not limited to initial industrial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill attendance records, and leadership training for industrial fire brigades shall be maintained for each industrial fire brigade member.

Compliance Statement Comnliance Basis Section (a) (4): Complies Section (a) (4): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL TAP-404,Training Documentation and Records ALL Chapter 3 Reouirement: (b) Training for Non-Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial fire brigade shall be trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade.

Compliance Statement Compliance Basis Section (b): Complies with Clarification Section (b): Guidance for non-industrial fire brigade members is found in FP-001.

HBRSEP LAR Rev 0 Page A-27

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements The procedure defines the actions needed to be taken by personnel discovering a fire, security personnel actions, and duty health physics contact actions.

Reference Document Doc Details FP-001 ,Fire Emergency ALL Chapter 3 Reauirement: (c)

• Drills. All of the following requirements shall apply.

(1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire brigade.

Compliance Statement Compliance Basis Section (c) (1): Complies Section (c) (1): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10.3.3.a Chapter 3 Requirement: (c)

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

These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire-fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario.

Compliance Statement Compliance Basis Section (c) (2): Complies Section (c) (2): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10 Chapter 3 Requirement: (c)

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

Comoliance Statement Comoliance Basis Section (c) (3): Complies Section (c) (3): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9,10.2 ChaPter 3 Requirement: (c)

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

Comollance Statement Comoliance Basis Section (c) (4): Complies Section (c) (4): No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL TAP-404,Training Documentation and Records ALL HBRSEP LAR Rev 0 Page A-28

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chanter 3 Reauirement: (c)

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

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

Reference:

3.4.4 Fire-Fighting Equipment.

Chapter 3 Reauirement: 3.4.4 Fire-Fighting Equipment.

Protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards.

Compliance Statement Compliance Basis Complies with Clarification 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."

HBRSEP has not committed to following any NFPA standards pertaining to firefighting equipment. Firefighting equipment is provided. A monthly inspection/inventory "of Fire Protection equipment and supplies located in the Fire Equipment Staging areas to meet the demands of the site Fire Brigade..." is conducted per OST-639.

Personnel dosimeters are issued in accordance with the plant radiation protection program and DOS-NGGC-0002. HP personnel, who provide fire brigade support, provide radiation monitoring equipment in accordance with FP-001.

HBRSEP LAR Rev 0 Page A-29

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details FAQ 06-0020,ldentification of "applicable NFPA standards" ALL OST-639,Fire Equipment Inventory (Monthly) ALL FP-001 ,Fire Emergency Section 3.7 OMM-002,Fire Protection Manual Section 3.27 DOS-NGGC-0002,Dosimetry Issuance ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5 Off-Site Fire Department Interface.

Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.1 Mutual Aid Agreement.

Chanter 3 Requirement: 3.4.5.1 Mutual Aid Agreement.

Off-site fire authorities shall be offered a plan for their interface during fires and related emergencies on site.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details PLP-007,Robinson Emergency Plan Attachment 6.2 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.2 Site-Specific Training.

Chapter 3 Requirement: 3.4.5.2* Site-Specific Training.

Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited to participate in a drill at least annually.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.7.1 HBRSEP LAR Rev 0 Page Ao30

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.5.3 Security and Radiation Protection.

Chapter 3 Requirement: 3.4.5.3* Security and Radiation Protection.

Plant security and radiation protection plans shall address off-site fire authority response.

Comoliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 8.10 PLP-007,Robinson Emergency Plan Table 5.3.2-1 Notes Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.4.6 Communications.

Chapter 3 ReQuirement: 3.4.6* Communications.

An effective emergency communications capability shall be provided for the industrial fire brigade.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details PLP-007,Robinson Emergency Plan Attachment 6.1 OST-639,Fire Equipment Inventory (Monthly) ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5 Water Supply Chanter 3 Requirement: N/A Compliance Statement Comeliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.1 [Water Supply Flow Code Requirements]

Chapter 3 Requirement: 3.5.1 A fire protection water supply of adequate reliability, quantity, and duration shall be provided by one of HBRSEP LAR Rev 0 Page A-31

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements the two following methods.

(a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies.

(b) Calculate the fire flow rate for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This fire flow rate shall be based on 500 gpm (1892.5 Llmin) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system (s) in the power block as determined in accordance with NFPA 13, Standard for the Installation of Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.

Comoliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with Section (b) of NFPA 805 Ch. 3 Section 3.5.1 as detailed in RNP-M/BMRK-1 011, RNP-M/MECH-1727, and RNP-M/MECH-1728.

Reference Document Doc Details RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Water Code Section 3012 Spray Fixed Systems RNP-M/MECH-1727,Hydraulic Analysis of the Hydrogen Seal Oil Water Section 5 Spray System RNP-M/MECH-1 728,Hydraulic Analysis of the Auxiliary &Start-Up Section 5 Transformer Water Spray System NLU-78-71 ,License Amendment 31 Section 4.3.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.5.2 [Water Supply Tank Code Requirements]

Chapter 3 Reauirement: 3.5.2*

The tanks shall be interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.

Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction from a large body of water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated.

Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service.

Compllance Statement Compliance Basis Complies with Clarification Fire water is obtained directly from Lake Robinson via a single intake structure.

Two physically separated automatic fire pumps are provided with separate suction lines.

As such, HBRSEP complies with Exception No. 1 HBRSEP LAR Rev 0 Page A-32

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details NLU-78-71 ,License Amendment 31 Section 4.3.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.3 [Water Supply Pump Code Requirements]

Chapter 3 Reouirement: 3.5.3*

Fire pumps, designed and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source.

Compliance Statement Comoliance Basis.

Complies via Engineering Evaluation HBRSEP complies with NFPA 20 as evaluated in RNP-M/BMRK-1012, RNP-M/MECH-1725, and RNP-M/MECH-1610.

Hydraulic analysis demonstrates the ability of one pump to provide required flow rate to the largest system.

Reference Document Doc Details RNP-M/MECH-1725,Evaluation of NFPA 13 Code Compliance Variances ALL RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Centrifugal Conclusions (Page 9) &Attachment 6-Code Section 33-Fire Pumps Page 4 of 62 FP-012,Fire Protection Systems Minimum Equipment and Compensatory Section 8.2 Actions RNP-M/MECH-1610,Hydraulic Analysis - Main Transformers ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.4 [Water Supply Pump Diversity and Redundancy]

Chapter 3 Requirement: 3.5.4 At least one diesel engine-driven fire pump or two more seismic Category I Class IE electric motor-driven fire pumps connected to redundant Class IE emergency power buses capable of providing 100 percent of the required flow rate and pressure shall be provided.

Compliance Statement Comoliance Basis Complies No Additional Clarification.

Reference Document Doc Details RNP-MIBMRK-1012,Code Compliance Evaluation NFPA 20 - Centrifugal Body of Calculation Fire Pumps FP-012,Fire Protection Systems Minimum Equipment and Compensatory Section 8.2 Actions HBRSEP LAR Rev 0 Page A-33

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.5.5 [Water Supply Pump Separation Requirements]

Chapter 3 Reauirement: 3.5.5 Each pump and its driver and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers.

Compliance Statement Compliance Basis Complies via Previous NRC Approval In a submittal dated 6/23/77, HBRSEP provided the following information:

"The Unit 2 fire protection water supply system meets the intent of the requirements (BTP ABCSB 9.5-1) either by itself or by virtue of cross-connection to the Unit 1 system, except that the Unit 2 supply pumps are not separated from each other or remaining pumps by three-hour rated fire walls. However, the various fire pumps are out-of-doors and separated by distance as well as intervening equipment."

In the SER dated 2/21/1980, the NRC stated:

"The staff does not agree with the licensee's contention that the arrangement of the propane storage tank and other equipment is satisfactory in relation to safety-related equipment on the intake structure for the following reasons....

Therefore, we will require the licensee to:

-Replace the propane engine with a diesel engine, or

-Replace the propane engine-driven fire pump and associated equipment to a location substantially remote from any safety-related equipment. "

The original propane-fueled engine driver on one of the two 100% capacity fire pumps was changed to a diesel-fueled driver per Modification M-445P to address concerns raised by the NRC over the propane storage location and arrangement.

Proposed modifications were evaluated and implemented per the SERs, where HBRSEP LAR Rev 0 Page A-34

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the basis for approval.

Reference Document Doc Details NLU-80-106,RFI and Requirements to Resolve Issues Concerning Fire Section 3.2.3 Protection NG-77-704,Fire Protection Program Review Question 15 M-445P,Fire Pump Engine Replacement & Propane Tank Relocation ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.6 [Water Supply Pump Start/Stop Requirements]

Chapter 3 Requirement: 3.5.6 Fire pumps shall be provided with automatic start and manual stop only.

Compliance Statement Compliance Basis Complies via Engineering Evaluation Fire pumps are provided with automatic start and manual stops, as detailed in the applicable portions of the NFPA 20 code compliance evaluation RNP-M/BMRK-1012.

Reference Document Doc Details RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Centrifugal Code Sections 515 (Attachment 6), Code Section 9-5 Fire Pumps (Attachment 7), and Code Section 9-5 (Attachment 8)

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.7 [Water Supply Pump Connection Requirements]

Chapter 3 ReQuirement: 3.5.7 Individual fire pump connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections.

Compliance Statement Compliance Basis Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:

"4.3.1.3 Fire Water Piping System The two fire pumps have a common discharge through a twelve inch underground main into a ten-inch underground fire water loop which HBRSEP LAR Rev 0 Page A-35

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements encircles the plant.

All yard fire hydrants, automatic water suppression systems and interior fire hose lines are supplied by the fire loop.

Sectionalizing valves of the post indicator type are provided on the fire loop to permit partial pipeline isolation without interruption of service to the entire system. The licensee will install isolation valves at the connection of the ten inch fire line from the Unit 1 fire loop to the Unit 2 fire loop, and provide separate headers for automatic sprinkler systems to be installed in the reactor auxiliary building.

The licensee will also provide barriers around all hydrants and post indicator valves to protect against vehicular damage.

Electrical supervision, to monitor the position of fire water system control valves, is not provided. A means of sealing these valves open will be provided, and this in combination with administrative controls and periodic inspections will be used to assure that valves are maintained open.

We find that, subject to the implementation of the above described modifications, the fire water piping systems satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable."

As seen in drawings HBR2-08255-Sheets 1 and 2, isolation valves were installed at the connection of the Unit 1 fire loop lines to the Unit 2 fire loop and separate headers for automatic sprinkler systems were installed in the reactor auxiliary building.

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the HBRSEP LAR Rev 0 Page A-36

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements basis for approval.

Reference Document Doc Details NLU-78-71 ,License Amendment 31 4.3.1.3 HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow Diagram ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.8 [Water Supply Pressure Maintenance Limitations]

Chapter 3 Reouirement: 3.5.8 A method of automatic pressure maintenance of the fire protection water system shall be provided independent of the fire pumps.

Compliance Statement Comoliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.2.7 RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Standpipes Code Section 661 and Hose Stations HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow Diagram ALL APP-044,Fire Alarm Console (FAC) C55, C58 Table B-1 NFPA 805 Ch.3 Transition Details Chafter 3

Reference:

3.5.9 [Water Supply Pump Operation Notification]

Chapter 3 Reauirement: 3.5.9 Means shall be provided to immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.3.4.1.c APP-044,Fire Alarm Console (FAC) C55, C58 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.10 [Water Supply Yard Main Code Requirements]

HBRSEP LAR Rev 0 Page A-37

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chaoter 3 Reauirement: 3.5.10 An underground yard fire main loop, designed and installed in accordance with NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish anticipated water requirements.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 24 as evaluated in RNP-M/BMRK-1013.

Reference Document Doc Details RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24 - Standard for ALL Outside Protection RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code Compliance Section 4.1 Variances Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.11 [Water Supply Yard Main Maintenance Issues]

Chapter 3 Requirement: 3.5.11 Means shall be provided to isolate portions of the yard fire main loop for maintenance or repair without simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations provided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems.

Compliance Statement Compliance Basis Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:

"4.3.1.3 Fire Water Piping System The two fire pumps have a common discharge through a twelve inch underground main into a ten-inch underground fire water loop which encircles the plant.

All yard fire hydrants, automatic water suppression systems and interior hose lines are supplied by the fire loop.

Sectionalizing valves of the post indicator type are provided on the fire loop to permit partial pipeline isolation without interruption of service to the entire system. The licensee will install isolation valves at the connection of the ten inch fire line from the Unit 1 fire loop to the Unit 2 fire loop, and provide separate headers for automatic sprinkler systems to be installed in the reactor auxiliary building.

HBRSEP LAR Rev 0 Page A-38

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements The licensee will also provide barriers around all fire hydrants and post indicator valves to protect against vehicular damage.

Electrical supervision, to monitor the position of fire water system control valves, is not provided. A means of sealing these valves open will be provided, and this in combination with administrative controls and periodic inspections will be used to assure that valves are maintained open.

We find that, subject to the implementation of the above described modifications, the fire water piping systems satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable."

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the basis for approval.

Reference Document Doc Details NLU-78-71 ,License Amendment 31 Section 4.3.1.3 HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow Diagram ALL HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL HBR2-08255 Sh. 6,Fire Protection System Flow Diagram ALL Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.5.12 [Water Supply Compatible Thread Connections]

Chapter 3 Reouirement: 3.5.12 Threads compatible with those used by local fire departments shall be 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.

Compliance Statement Compliance Basis Complies No Additional Clarification HBRSEP LAR Rev 0 Page A-39

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.2.5 Table B-1 NFPA 805 Ch.3 Transition Details Chagter 3

Reference:

3.5.13 [Water Supply Header Options]

Chapter 3 Requirement: 3.5.13 Headers fed from each end shall be permitted inside buildings to supply both sprinkler and standpipe systems, provided steel piping and fittings meeting the requirements of ANSI B31.1, Code for Power Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems where such headers are part of the seismically analyzed hose standpipe system. Where provided, such headers shall be considered an extension of the yard main system. Each sprinkler and standpipe system shall be equipped with an outside screw and yoke, gate valve or other approved shutoff valve.

Compliance Statement Compliance Basis N/A No headers at HBRSEP are fed from each end.

Reference Document Doc Details HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.14 [Water Supply Control Valve Supervision]

Chapter 3 Requirement: 3.5.14*

All fire protection water supply and fire suppression system control valves shall be under a periodic inspection program and shall be supervised by one of the following methods.

(a) Electrical supervision with audible and visual signals in the main control room or other suitable constantly attended location.

Compliance Statement Compliance Basis Complies HBRSEP complies with Section 3.5.14 by a combination of (b) & (c).

Reference Document Doc Details OST-602,Unit No. 2 Fire Water System Flowpath Verification (Monthly) and ALL Valve Cycling (Annual)

OST-652,Unit 2 Containment Fire Water System Valves ALL NLU-78-71,License Amendment 31 Section 4.3.1.3 Chapter 3 Requirement: (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel.

Compliance Statement Compliance Basis Complies HBRSEP complies with Section 3.5.14 by a combination of (b) & (c).

HBRSEP LAR Rev 0 Page A-40

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details OST-602,Unit No. 2 Fire Water System Flowpath Verification (Monthly) and ALL Valve Cycling (Annual)

OST-652,Unit 2 Containment Fire Water System Valves ALL NLU-78-71 ,License Amendment 31 Section 4.3.1.3 Chapter 3 Reauirement: (c) Sealing valves in their normal positions. This option shall be utilized only where valves are located within fenced areas or under the direct control of the owner/operator.

Compliance Statement Compliance Basis Complies HBRSEP complies with Section 3.5.14 by a combination of (b) & (c).

Reference Document Doc Details OST-602,Unit No. 2 Fire Water System Flowpath Verification (Monthly) and ALL Valve Cycling (Annual)

OST-652,Unit 2 Containment Fire Water System Valves ALL NLU-78-71 ,License Amendment 31 Section 4.3.1.3 Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.5.15 [Water Supply Hydrant Code Requirements]

Chapter 3 ReQuirement: 3.5.15 Hydrants shall be installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be provided at intervals of not more than 1000 ft (305 m) along the yard main system.

Exception: Mobile means of providing hose and associated equipment, such as hose carts or trucks, shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to the equipment supplied by three hose houses.

Compliance Statement Compliance Basis Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC APPROVAL: In the SER dated 2/28/78, Complies via Engineering Evaluation the NRC stated:

"Five yard fire hydrants are provided at approximately 250 foot intervals around the exterior of the plant except at the north end of the plant, where the distance between hydrants is somewhat larger. A hose house located near each fire hydrant contains 2-1/2 inch diameter fire hose and other manual firefighting tools. A sixth hose house is centrally located between the reactor and Turbine Building, a seventh is located at the intake structure, and one hose house is located outside the HBRSEP LAR Rev 0 Page A-41

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements Unit 2 fence east of the auxiliary building.

Standard fire hose threads are used on all fire protection equipment, and the threads are compatible with those used by the local public fire departments.

We find that, subject to the implementation of the above described modifications, the fire water piping systems satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable."

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the basis for approval.

COMPLIES VIA ENGINEERING EVALUATION: HBRSEP complies with the applicable portions of NFPA 24 as detailed in RNP-M/BMRK-1013.

Reference Document Doc Details NLU-78-71,License Amendment 31 Section 4.3.1.3 RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24 - Standard for ALL Outside Protection RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code Compliance ALL Variances Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.5.16 [Water Supply Dedicated Limits]

Chapter 3 Requirement: 3.5.16*

The fire protection water supply system shall be dedicated for fire protection use only.

Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.

Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.

HBRSEP LAR Rev 0 Page A-42

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Compliance Statement Compliance Basis License Amendment Required NRC approval is being requested in Attachment L for the use of the fire protection water supply system for purposes other than fire protection.

Reference Document Doc Details HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow Diagram ALL HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL HBR2-08255 Sh 3,Fire Protection System Containment Flow Diagram ALL HBR2-08255 Sh. 6,Fire Protection System Flow Diagram ALL OMM-002,Fire Protection Manual Section 8.15 AOP-014,Component Cooling Water System Malfunction ALL AOP-022,Loss of Service Water ALL EDMG-001,Extreme Damage Event Early Actions and Response ALL Determination Criteria EDMG-002,Refueling Water Storage Tank (RWST) ALL EDMG-003,Condensate Storage Tank (CST) ALL EDMG-005,Containment Vessel (CV) ALL EDMG-011 ,Spent Fuel Pool Casualty ALL EDMG-012,Core Cooling Using Alternate Water Source ALL EDMG-013,Airbome Release Scrubbing ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6 Standpipe and Hose Stations.

Chapter 3 Requirement: N/A Compliance Statement Compliance Basis N/A N/A Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.1 [Standpipe and Hose Station Code Requirements]

Chapter 3 Requirement: 3.6.1 For all power block buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.

Compliance Statement Compliance Basis Complies via Engineering Evaluation COMPLIES VIA ENGINEERING EVALUATION: HBRSEP complies with Complies via Previous NRC Approval NFPA 14 as detailed in RNP-M/BMRK-1010 and RNP-M/MECH-1709.

HBRSEP LAR Rev 0 Page A-43

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements COMPLIES VIA PREVIOUS NRC APPROVAL: HBRSEP has Class II standpipes in lieu of Class II1.In the License Amendment dated 2/28/78, the NRC stated:

"4.3.1.4 Interior Fire Hose Stations A total of 24 interior hose stations, each presently equipped with 50 feet of 1-1/2 inch diameter hose, have been provided throughout all portions of the plant except containment. There are presently several safety-related areas containing combustible materials that are beyond the reach of the existing hose lines. The licensee will provide additional hose stations or additional lengths of hose at existing stations so that sufficient hose reach is provided to protect all the areas of the auxiliary building.

Hose racks originally used for unlined hose are being used to store rubber lined hose. The licensee has committed to replace these with suitable hose reels or hose racks designed and sized for lined hose.

The nozzles on the interior hose lines are 1-1/2" spray nozzles. In areas with electrical hazards, "electrically safe" hose nozzles have been provided on the hose station nearest these areas.

We find that, subject to the implementation of the above described modifications, the interior fire hose stations satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable."

Proposed modifications were evaluated and implemented per the SERs, where applicable, to fulfill the intent of this requirements.

There have been no plant modifications or other changes that would invalidate the basis for approval.

Reference Document Doc Details RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Standpipes ALL and Hose Stations RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code Compliance ALL Variances HBRSEP LAR Rev 0 Page A-44

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements NLU-78-71 ,License Amendment 31 4.3.1.4 Table B-1 NFPA 805 Ch.3 Transition Details Chaotar 3

Reference:

3.6.2 (Standpipe and Hose Station Capability Limitations]

Chapter 3 Requirement: 3.6.2 A capability shall be provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel.

Compliance Statement Comoliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 805 requirement 3.6.2 as detailed in the applicable portions RNP-M/BMRK-1010.

Reference Document Doc Details RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Standpipes ALL and Hose Stations Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.6.3 [Standpipe and Hose Station Nozzle Restrictions]

Chanter 3 Reouirement: 3.6.3 The proper type of hose nozzle to be supplied to each power block area shall be based on the area fire hazards. The usual combination spray/straight stream nozzle shall not be used in areas where the straight stream can cause unacceptable damage or present an electrical hazard to fire-fighting personnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 805 requirement 3.6.3 as detailed in the applicable portions of RNP-M/BMRK-1 010.

Reference Document Doc Details RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Standpipes Code Section 451 &452 and Hose Stations NLU-78-71 ,License Amendment 31 Section 4.3.1.4 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.4 [Standpipe and Hose Station Earthquake Provisions]

HBRSEP LAR Rev 0 Page A-45

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements Chapter 3 Reauirement: 3.6.4 Provisions shall be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE).

Compliance Statement Compliance Basis Complies with Clarification Seismic standpipes are not an original commitment for HBRSEP.

The Federal Register notice that promulgated adoption of NFPA 805 makes the following statement:

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

There have been no plant modifications or other changes that would invalidate the basis for approval.

Reference Document Doc Details 69 FR 33356,Final Rule - NFPA 805 Page 33544 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.6.5 (Standpipe and Hose Station Seismic Connection Limitations]

Chapter 3 Reauirement: 3.6.5 Where the seismic required hose stations are cross-connected to essential seismic non-fire protection water supply systems, the fire flow shall not degrade the essential water system requirement.

Compliance Statement Compliance Basis N/A HBRSEP is not committed to having seismic standpipes. See the Compliance Section for 3.6.4 HBRSEP LAR Rev 0 Page A-46

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chap~ter 3

Reference:

3.7 Fire Extinguishers.

Chaster 3 Reauirement: 3.7 Fire Extinguishers.

Where provided, fire extinguishers of the appropriate number, size, and type shall be provided in accordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 10 as evaluated in RNP-M/BMRK-1001.

Reference Document Doc Details RNP-M/BMRK-1001 ,Code Compliance Evaluation NFPA 10 Portable Fire ALL Extinguishers Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8 Fire Alarm and Detection Systems.

Chapter 3 Reouirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.1 Fire Alarm.

Chapter 3 Requirement: 3.8.1 Fire Alarm.

Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code.

Alarm annunciation shall allow the proprietary alarm system to transmit fire-related alarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnel assigned to the proprietary alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted:

(1) Actuation of any fire detection device Compliance Statement Compliance Basis (Main Header): Complies via Engineering (Main Header): HBRSEP complies with Evaluation NFPA 72 as evaluated in the applicable portions of RNP-M/BMRK-1014, 1005, Section (1): Complies 1006.

Section (1): No Additional Clarification HBRSEP LAR Rev 0 Page A-47

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National Fire ALL Alarm Code RNP-M/BMRK-1005,Code Compliance Evaluation NFPA 72D ALL RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E ALL RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance Variances ALL APP-044,Fire Alarm Console (FAC) ALL UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-14 Chapter 3 Requirement: (2) Actuation of any fixed fire suppression system Compliance Statement Compliance Basis Section (2): Complies Section (2): No Additional Clarification Reference Document Doc Details APP-044,Fire Alarm Console (FAC) ALL Chapter 3 Requirement: (3) Actuation of any manual fire alarm station Comoliance Statement Compliance Basis Section (3): Complies Section (3): No Additional Clarification Reference Document Doc Details APP-044,Fire Alarm Console (FAC) ALL Chapter 3 Reguirement: (4) Starting of any fire pump Compliance Statement Compliance Basis Section (4): Complies Section (4): No Additional Clarification Reference Document Doc Details APP-044,Fire Alarm Console (FAC) ALL Chapter 3 Reouirement: (5) Actuation of any fire protection supervisory device Compliance Statement Compliance Basis Section (5): Complies Section (5): No Additional Clarification Reference Document Doc Details APP-044,Fire Alarm Console (FAC) ALL Chapter 3 Requirement: (6) Indication of alarm system trouble condition Comoliance Statement Compliance Basis Section (6): Complies Section (6): No Additional Clarification Reference Document Doc Details APP-044,Fire Alarm Console (FAC) ALL HBRSEP LAR Rev 0 Page A-48

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.8.1.1 [Fire Alarm Communication Requirements]

Chapter 3 Reauirement: 3.8.1.1 Means shall be provided to allow a person observing a fire at any location in the plant to quickly and reliably communicate to the control room or other suitable constantly attended location.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details FP-001 ,Fire Emergency Section 5.2 UFSARHBR 2 Updated Final Safety Analysis Report (FSAR) Sections 9.5.2 Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.8.1.2 [Fire Alarm Prompt Notification Limits]

Chaoter 3 Reauirement: 3.8.1.2 Means shall be provided to promptly notify the following of any fire emergency in such a way as to allow them to determine an appropriate course of action:

(1) General site population in all occupied areas.

Comoliance Statement Compliance Basis Section (1): Complies Section (1): No Additional Clarification Reference Document Doc Details AOP-041 ,Response to Fire Event Section 2 Chapter 3 Reauirement: (2) Members of the industrial fire brigade and other groups supporting fire emergency response Compliance Statement Compliance Basis Section (2): Complies Section (2): No Additional Clarification Reference Document Doc Details AOP-041 ,Response to Fire Event Section 2 Chapter 3 Reauirement: (3) Off-site fire emergency response agencies. Two independent means shall be available (e.g.,

telephone and radio) for notification of off-site emergency services Compliance Statement Comoliance Basis Section (3): Complies Section (3): No Additional Clarification Reference Document Doc Details PLP-007,Robinson Emergency Plan Attachment 6.1 Table B-1 NFPA 805 Ch.3 Transition Details HBRSEP LAR Rev 0 Page A-49

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chapter 3

Reference:

3.8.2 Detection.

Chapter 3 Requirement: 3.8.2 Detection.

Ifautomatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 72 as evaluated in the applicable portions of RNP-M/BMRK-1014, 1005,1006.

Reference Document Doc Details RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National Fire ALL Alarm Code RNP-M/BMRK-1005,Code Compliance Evaluation NFPA 72D ALL RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E ALL RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance Variances ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9 Automatic and Manual Water-Based Fire Suppression Systems.

Chapter 3 Reauirement: N/A Compliance Statement Compliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.9.1 [Fire Suppression System Code Requirements]

Chapter 3 Reauirement: 3.9.1

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

(1) NFPA 13, Standard for the Installation of Sprinkler Systems Compliance Statement Compliance Basis Section (1): Complies via Engineering Section (1): HBRSEP complies with NFPA Evaluation 13 as evaluated in the applicable portions of RNP-M/BMRK-1009.

Reference Document Doc Details RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sprinkler ALL HBRSEP LAR Rev 0 Page A-50

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Systems RNP-M/MECH-1725,Evaluation of NFPA 13 Code Compliance Variances ALL Chaoter 3 Reauirement: (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection Comaliance Statement Compliance Basis Section (2): Complies via Engineering Section (2): HBRSEP complies with NFPA Evaluation 15 as evaluated in the applicable portions of RNP-M/BMRK-1011.

Reference Document Doc Details RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Water ALL Spray Fixed Systems RNP-M/MECH-1 726,Evaluation of NFPA 15 Code Compliance Variances ALL RNP-M/MECH-1 727,Hydraulic Analysis of the Hydrogen Seal Oil Water ALL Spray System RNP-M/MECH-1 728,Hydraulic Analysis of the Auxiliary & Start-Up ALL Transformer Water Spray System Chaoter 3 Reauirement: (3) NFPA 750, Standard on Water Mist Fire Protection Systems Comoliance Statement Compliance Basis Section (3): N/A Section (3): N/A - No Water Mist Fire Protection Systems are installed at HBRSEP.

Chapter 3 Reauirement: (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems Compliance Statement Compliance Basis Section (4): N/A Section (4): N/A - No Foam-Water Sprinkler or Foam-Water Spray Systems are installed at HBRSEP.

Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.2 [Fire Suppression System Flow Alarm]

Chapter 3 Reauirement: 3.9.2 Each system shall be equipped with a water flow alarm.

Compliance Statement Compliance Basis Complies with Clarification COMPLIES WITH CLARIFICATION:

Some automatic water-based fire Complies via Engineering Evaluation suppression systems do not have water flow alarms. These systems are not required to have water flow alarms per NFPA 13, which only requires water flow alarms to be provided on sprinkler systems having more than 20 sprinklers.

HBRSEP LAR Rev 0 Page A-51

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program &Design Elements COMPLIES VIA ENGINEERING EVALUATION: Per RNP-M/BMRK-1009, "Code Compliance Evaluation NFPA 13 - Standard for Installation For Sprinkler Systems" & RNP-M/BMRK-1011, "Code Compliance Evaluation NFPA 15 - Water Spray Fixed Systems",

all requirements from NFPA 13 & NFPA 15 for water flow alarms were met.

Reference Document Doc Details RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sprinkler Code Sections 3-16.2, 3-17.2 Systems RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Water Code Section 2124 Spray Fixed Systems Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.3 [Fire Suppression System Alarm Locations]

Chapter 3 Reauirement: 3.9.3 All alarms from fire suppression systems shall annunciate in the control room or other suitable constantly attended location.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 805 requirement 3.9.3 as detailed in the applicable portions of RNP-M/BMRK-1009.

Reference Document Doc Details RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sprinkler Code Sections 5-3.5.2, 3-17.3.3, 3-17.6.2 Systems RNP-MIBMRK-1011,Code Compliance Evaluation for NFPA 15, Water Code Section 8041 Spray Fixed Systems APP-044,Fire Alarm Console (FAC) ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.4 [Fire Suppression System Diesel Pump Sprinkler Protection]

Chapter 3 Requirement: 3.9.4 Diesel-drven fire pumps shall be protected by automatic sprinklers.

Comoliance Statement Compliance Basis Complies with Clarification The Diesel -driven fire pump is located outdoors and is separated from other important equipment.

HBRSEP LAR Rev 0 Page A-52

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60 Sheet 45 Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.9.5 [Fire Suppression System Shutoff Controls]

Chapter 3 Reauirement: 3.9.5 Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 805 requirement 3.9.5 as detailed in the applicable portions of RNP-M/BMRK-1009 and RNP-M/BMRK-1011.

Reference Document Doc Details RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sprinkler Sections 3-13.1.1, 3-14.1.1 Systems RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15, Water Section 2080 Spray Fixed Systems Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.9.6 [Fire Suppression System Valve Supervision]

Chapter 3 Reauirement: 3.9.6 All valves controlling water-based fire suppression systems required to meet the performance or deterministic requirements of Chapter 4 shall be supervised as described in 3.5.14.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OST-602,Unit No. 2 Fire Water System Flowpath Verification (Monthly) and ALL Valve Cycling (Annual)

OST-652,Unit 2 Containment Fire Water System Valves ALL NLU-78-71 ,License Amendment 31 Section 4.3.1.3 Table B-1 NFPA 805 Ch.3 Transition Details Chaotar 3

Reference:

3.10 Gaseous Fire Suppression Systems.

Chapter 3 Reauirement: N/A Comoliance Statement Comollance Basis HBRSEP LAR Rev 0 Page A-53

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements N/A N/A - General statement; No technical requirements.

Table B-1 NEPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.10.1 [Gaseous Suppression System Code Requirements]

Chapter 3 Reauirement: 3.10.1 If an automatic total flooding and local application gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following applicable NFPA codes:

(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems Compliance Statement Comoliance Basis Section (1): Complies via Engineering Section (1): HBRSEP complies with NFPA Evaluation 12 as evaluated in the applicable portions of RNP-M/BMRK-1007.

Reference Document Doc Details RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon ALL Dioxide Extinguishing Systems RNP-MIMECH-1708,Evaluation of NFPA 12 Code Compliance Variances ALL HBR2-11992 SHOO001,EDG HIGH PRESSURE C02 FIRE ALL EXTINGUISHING SYSTEM HBR2-11992 Sh 02,EDG High Pressure C02 Fire Extinguishing System ALL HBR2-11992 Sh 03,EDG High Pressure C02 Fire Extinguishing System ALL Chapter 3 Reauirement: (2) NFPA 12A, Standard on Halon 1301. Fire Extinguishing Systems Compliance Statement Compliance Basis Section (2): Complies via Engineering Section (2): HBRSEP complies with NFPA Evaluation 12A as evaluated in the applicable portions of RNP-M/BMRK-1008.

Reference Document Doc Details RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, Halon ALL 1301 Systems Chapter 3 Requirement: (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems Comoliance Statement Compliance Basis Section (3): N/A Section (3): Clean Agent Fire Extinguishing Systems are not utilized at HBRSEP.

Table B-1 NFPA 805 Ch.3 Transition Details HBRSEP LAR Rev 0 Page A-54

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Chapter 3

Reference:

3.10.2 [Gaseous Suppression System Alarm Location]

Chapter 3 Requirement: 3.10.2 Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other constantly attended location identified.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon Section 1452 & 1-8.5 Dioxide Extinguishing Systems RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, Halon Section 1-8.4 1301 Systems APP-044,Fire Alarm Console (FAC) ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.3 [Gaseous Suppression System Ventilation Limitations]

Chaoter 3 Reauirement: 3.10.3 Ventilation system design shall take into account prevention from over-pressurization during agent injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants.

Compliance Statement Compliance Basis Complies via Engineering Evaluation HBRSEP complies with NFPA 805 requirement 3.10.3 as detailed in the applicable portions of RNP-M/BMRK-1007 and RNP-M/BMRK-1008.

Reference Document Doc Details RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon ALL Dioxide Extinguishing Systems RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, Halon ALL 1301 Systems RNP-M/MECH-1708,Evaluation of NFPA 12 Code Compliance Variances ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.4 [Gaseous Suppression System Single Failure Limits]

Chapter 3 Reouirement: 3.10.4*

In any area required to be protected by both primary and backup gaseous fire suppression systems, a single active failure or a crack in any pipe in the fire suppression system shall not impair both the primary and backup fire suppression capability.

Compliance Statement Compliance Basis N/A No areas at HBRSEP are required to be protected by both primary and backup HBRSEP LAR Rev 0 Page A-55

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements gaseous fire suppression systems.

Table B-1 NFPA 805 Ch.3 Transition Details Chaoter 3

Reference:

3.10.5 [Gaseous Suppression System Disarming Controls]

Chapter 3 Reauirement: 3.10.5 Provisions for locally disarming automatic gaseous suppression systems shall be secured and under strict administrative control.

Compliance Statement Comoaliance Basis Complies No Additional Clarification Reference Document Doc Details OMM-002,Fire Protection Manual Section 8.13.6 OP-809,Diesel Generator Carbon Dioxide Suppression System ALL OPS-NGGC-1308,Plant Status Control ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.6 [Gaseous Suppression System C02 Limitations]

Chaoter 3 Reauirement: 3.10.6*

Total flooding carbon dioxide systems shall not be used in normally occupied areas.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details HBR2-9717,Fire Area/Zone Locations ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.7 [Gaseous Suppression System C02 Warnings]

ChaOter 3 Reauirement: 3.10.7 Automatic total flooding carbon dioxide systems shall be equipped with an audible pre-discharge alarm and discharge delay sufficient to permit egress of personnel. The carbon dioxide system shall be provided with an odorizer.

Compliance Statement Compliance Basis Complies with Clarification See proposed modification pertinent to NFPA 805 Chapter 3, Section 3.10.7 compliance in Attachment "S", Table S-2 of the Transition Report.

HBRSEP LAR Rev 0 Page A-56

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon Code Sections 122 & 1-6.2 Dioxide Extinguishing Systems Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.8 [Gaseous Suppression System C02 Required Disarming]

Chapter 3 ReauIrement: 3.10.8 Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during work in the protected space.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details OP-805,Carbon Dioxide Suppression System Section 8.3 OP-809,Diesel Generator Carbon Dioxide Suppression System Section 8.3 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.9 [Gaseous Suppression System Cooling Considerations]

Chaoter 3 Reouirement: 3.10.9 The possibility of secondary thermal shock (cooling) damage shall be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide.

Compliance Statement Comoliance Basis Complies No Additional Clarification Reference Document Doc Details UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1 BTP APCSB 9.5-1 ,Guideline for Fire Protection for Nuclear Power Plants Appendix A Section E.4 & E.5 Docketed Prior to July 1, 1976 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.10.10 [Gaseous Suppression System Decomposition Issues]

Chapter 3 Requirement: 3.10.10 Particular attention shall be given to corrosive characteristics of agent decomposition products on safety systems.

Compliance Statement Compliance Basis Complies No Additional Clarification HBRSEP LAR Rev 0 Page A-57

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Reference Document Doc Details BTP APCSB 9.5-1,Guideline for Fire Protection for Nuclear Power Plants Appendix A Section E.4 & E.5 Docketed Prior to July 1, 1976 UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11 Passive Fire Protection Features Chapter 3 Reauirement: 3.11 Passive Fire Protection Features.

This section shall be used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.

Compliance Statement Comoliance Basis N/A N/A - General statement; No technical requirements.

Table B-1 NFPA 805 Ch.3 Transition Details Chanter 3

Reference:

3.11.1 Building Separation.

Chapter 3 Reauirement: 3.11.1 Building Separation.

Each major building within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, Recommended Practice for Protection of Buildings from Exterior Fire Exposures.

Exception: Where a performance-based analysis determines the adequacy of building separation, the requirements of 3.11.1 shall not apply.

Comollance Statement Comoliance Basis Complies COMPLIES: No Additional Clarification Complies via Engineering Evaluation COMPLIES VIA ENGINEERING EVALUATION: Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the plant.

Reference Document Doc Details RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 1 HBRSEP LAR Rev 0 Page A-58

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements RNP-M/MECH-1 673,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-M/MECH-1674.Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 3 RNP-M/MECH-1 675,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-M/MECH-1 676,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-M/MECH-1 679,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1681 ,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1685,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1 691 ,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 23 RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 NLU-78-71 ,License Amendment 31 Sections 4.11 & 4.14 HBRSEP LAR Rev 0 Page A-59

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements FPP-RNP-900, Fire Hazards Analysis ALL HBR2-9717,Fire Area/Zone Locations ALL Table B-i NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.2 Fire Barriers.

Chapter 3 Requirement: 3.11.2 Fire Barriers.

Fire barriers required by Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be designed and installed to meet the specific fire resistance rating using assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Fire Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials.

Comoliance Statement Compliance Basis Complies COMPLIES: No Additional Clarification Complies via Engineering Evaluation COMPLIES VIA ENGINEERING EVALUATION: Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the plant.

Reference Document Doc Details RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 1 RNP-M/MECH-1673, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-M/MECH-1 674,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 3 RNP-MIMECH-1675 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-MIMECH-1 676,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1 677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-MIMECH-1679,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 HBRSEP LAR Rev 0 Page A-60

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements RNP-M/MECH-1683.Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1 684,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1 685,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 RNP-M/MECH-1 689,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1690, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1691 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-M/MECH-1 692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 23 RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 NLU-78-71 ,License Amendment 31 Sections 4.11 & 4.14 FP-014,Control of Fire Barrier Penetrations ALL RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration Seals ALL EE-87-0166,Evaluation Of Concrete Brick "Rubble" Fire Barrier Penetration ALL Seals ESR-97-405,Evaluation of Pyrocrete Fire Barrier Designs ALL EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting ALL ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL FPP-RNP-900,Fire Hazards Analysis ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.3 Fire Barrier Penetrations.

Chapter 3 ReQuirement: 3.11.3" Fire Barrier Penetrations.

Penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by Chapter 4. (See 3.11.3.4 for HBRSEP LAR Rev 0 Page A-61

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable:

(1) NFPA 80, Standard for Fire Doors and Fire Windows.

Compliance Statement Compliance Basis Section (1): Section (1):

Complies via Engineering Evaluation Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the plant.

HBRSEP complies with NFPA 80 as evaluated in the applicable portions of RNP-M/BMRK-1003.

Reference Document Doc Details RNP-M/MECH-1 672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 1 RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-M/MECH-1674, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 3 RNP-M/MECH-1675, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-M/MECH-1676, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-M/MECH-1679, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1685,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1 686,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 HBRSEP LAR Rev 0 Page A-62

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 RNP-MIMECH-1689,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1 691 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 23 RNP-MIMECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80 ALL STANDARD FOR FIRE DOORS AND WINDOWS UFSARHBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1A-7 RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration Seals ALL RNP-M/MECH-1670,Evaluation Of Concrete Hatch Covers ALL EE-90-0104,Generic Evaluation Of HVAC Fire Damper And Fire Door Installation Discrepancies Chapter 3 Reauirement: (2) NFPA 90A, Standard for the Installation of Air-Conditioning and Ventilating Systems.

Compliance Statement Compliance Basis Section (2): Complies via Engineering Section (2):

Evaluation Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the plant.

HBRSEP complies with NFPA 90A as evaluated in the applicable portions of RNP-M/BMRK-1004.

Reference Document Doc Details RNP-M/MECH-1 672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 1 RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-MIMECH-1674,Evaluation of Non-Standard Fire Barrier Penetration ALL HBRSEP LAR Rev 0 Page A-63

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Seals in Fire Zone 3 RNP-M/MECH-1 675,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-M/MECH-1 676,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1 677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-M/MECH-1679,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1 681 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 RNP-M/MECH-1683,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1685, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1686, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 RNP-M/MECH-1689 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1 691 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 23 RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR NFPA 90A ALL 1976 & 1985 EDITIONS AIR CONDITIONING UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1A-7 HBRSEP LAR Rev 0 Page A-64

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration Seals ALL ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL EE-90-0104,Generc Evaluation Of HVAC Fire Damper And Fire Door ALL Installation Discrepancies Chapter 3 Reauirement: (3) NFPA 101, Life Safety Code Exception: Where fire area boundaries are not wall-to-wall, floor-to-ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall be required to assess the adequacy of fire barrier forming the fire boundary to determine if the barrier will withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to be protected by other means as acceptable to the AHJ.

Compliance Statement ComPliance Basis Section (3): Complies with Clarification Section (3): HBRSEP complies with clarification with NFPA 101. HBRSEP complies with NFPA 101 with regards to fire rated door assemblies since NFPA 101, Section 8.3.3.1 refers to NFPA 80, which is evaluated in RNP-M/BMRK-1003. HBRSEP complies with NFPA 101 with regards to rated fire dampers since NFPA 101, Section 9.2.1 refers to NFPA 90A, which is evaluated in RNP-M/BMRK-1004 Reference Document Doc Details RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80 ALL STANDARD FOR FIRE DOORS AND WINDOWS RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR NFPA 90A ALL 1976 & 1985 EDITIONS AIR CONDITIONING NFPA 101,Life Safety Code, 2009 Edition Sections 8.3.3.1 & 9.2.1 Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.4 Through Penetration Fire Stops.

Chapter 3 Requirement: 3.11.4* Through Penetration Fire Stops.

Through penetration fire stops for penetrations such as pipes, conduits, bus ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall be protected as follows.

(a) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.

Compliance Statement Compliance Basis Section (a): Section (a):

Complies COMPLIES: No Additional Clarification HBRSEP LAR Rev 0 Page A-65

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Complies via Engineering Evaluation COMPLIES VIA ENGINEERING EVALUATION: Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the plant.

Engineering evaluations were developed to analyze the acceptability of the typical penetration seal designs utilized at HBRSEP.

Reference Document Doc Details RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone I RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 3 RNP-M/MECH-1675,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-MIMECH-1676,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-MIMECH-1679, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1681,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 RNP-M/MECH-1 683,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1684,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1685, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1686, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-M/MECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 HBRSEP LAR Rev 0 Page A-66

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1691 ,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-M/MECH-1692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 23 RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 HBR2-09716,Fire Barrier Penetrations ALL GID/R8703810014,Design Basis Document; Fire Barrier System Sections 4.1.3 & 4.1.4 FP-0 14,Control of Fire Barrier Penetrations ALL RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration Seals ALL EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting ALL ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL EE-90-0025,Past And Present Operability Of Steam Generator Blowdown ALL Line Penetration Seals (Penetrations CP-2674 And CP-5612)

EE-93-0043,Evaluation Of Temporary Fire Barrier Penetration Seals ALL Between Fire Zone 11 And 24 ESR-98-0221,Penetration Seals Containing Copper Piping and Tubing ALL RNP-M/MECH-1671,Evaluation Of Large Bore Piping Penetrations ALL Chapter 3 Reauirement: (b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and shall be permitted to be installed on either side of the barrier in a location that is as close to the barrier as possible.

Exception: Openings inside conduit 4 in. (10.2 cm) or less in diameter shall be sealed at the fire barrier with a fire-rated internal seal unless the conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot gas seal in this application.

Compliance Statement Compliance Basis Section (b): Complies Section (b):

COMPLIES: No Additional Clarification Complies via Engineering Evaluation COMPLIES VIA ENGINEERING EVALUATION: Design and installation deviations pertaining to passive fire protection features are evaluated on a fire zone basis at RNP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages as installed in the HBRSEP LAR Rev 0 Page A-67

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements plant.

Reference Document Doc Details RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 1 RNP-M/MECH-1673,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 2 RNP-M/MECH-1674,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 3 RNP-M/MECH-1675,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 4 RNP-M/MECH-1676,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 5 RNP-M/MECH-1677,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 6 RNP-M/MECH-1678,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 7 RNP-M/MECH-1679,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 8 RNP-M/MECH-1680,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 9 RNP-M/MECH-1 681 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 10 RNP-M/MECH-1682, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 11 RNP-M/MECH-1683, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 12 RNP-M/MECH-1684, Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 13 RNP-M/MECH-1685 Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 14 RNP-M/MECH-1686,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 15 RNP-M/MECH-1687,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 16 RNP-MIMECH-1688,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 17 RNP-M/MECH-1689,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 18 RNP-M/MECH-1690,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 19 RNP-M/MECH-1691 ,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 20 RNP-MIMECH-1692,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 21 RNP-M/MECH-1693,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 22 RNP-M/MECH-1694,Evaluation of Non-Standard Fire Barrier Penetration ALL HBRSEP LAR Rev 0 Page A-68

Attachment A - NEI 04-02 Table B-1 Transition of Fundamental Duke Energy Fire Protection Program & Design Elements Seals in Fire Zone 23 RNP-M/MECH-1695,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 24 RNP-M/MECH-1696,Evaluation of Non-Standard Fire Barrier Penetration ALL Seals in Fire Zone 27 GID/R8703810014,Design Basis Document; Fire Barrier System Section 2.2.2.1, Table A.4.0-1 CTL# CRE093-4324,Conduit Fire Test of One Hundred One Electrical ALL Conduit Penetrations RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration Seals ALL ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL ESR-94-0930,Fire Barriers ALL ESR-94-11 03,Evaluation for 12" of Dow Corning Foam For Conduit Seals ALL NED-B/MECH-t001,Fire Resistance of Capped Conduits ALL Table B-1 NFPA 805 Ch.3 Transition Details Chapter 3

Reference:

3.11.5 Electrical Raceway Fire Barrier Systems (ERFBS).

Chapter 3 Reauirement: 3.11.5" Electrical Raceway Fire Barrier Systems (ERFBS).

ERFBS required by Chapter 4 shall be capable of resisting the fire effects of the hazards in the area.

ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area." The ERFBS needs to adequately address the design requirements and limitations of supports and intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill, and type shall be demonstrated.

Exception No. 1: When the temperatures inside the fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Training Within the Same Fire Area," Supplement 1, functionality of the cable at these elevated temperatures shall be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure.

Exception No. 2: ERFBS systems employed prior to the issuance of Generic Letter 86-10, Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance.

Compliance Statement Compliance Basis Complies No Additional Clarification Reference Document Doc Details GID/R87038/0014,Design Basis Document; Fire Barrier System Section 4.4 Table B-1 NFPA 805 Ch.3 Transition Details HBRSEP LAR Rev 0 Page A-69

Duke Energy Attachment B - NEI 04-02 Table B-2 Nuclear Safety Capability Assessment Methodology Review B. NEI 04-02 Table B Nuclear Safety Capability Assessment -

Methodology Review 90 Pages Attached Page B-I HBRSEP LAR Rev 00 LAR Rev Page B-1

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3 Deterministic This section discusses a generic deterministic methodology and criteria that licensees can use to Methodology perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific analysis approved by NRC is reflected in the plant's licensing basis. The methodology described in this section is also an acceptable method of performing a post-fire safe shutdown analysis. This methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used.

Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships. This guidance document, however, does not require the use of a computer database oriented approach.

The requirements of Appendix R Sections Ill.G.1, III.G.2 and III.G.3 apply to equipment and cables required for achieving and maintaining safe shutdown in any fire area. Although equipment and cables for fire detection and suppression systems, communications systems and 8-hour emergency lighting systems are important features, this guidance document does not address them.

Additional information is provided in Appendix B to this document.

Ao1licabilitv Comments Applicable Alignment Statement Alignment Basis Statement Aligns Robinson Nuclear Plant's (HBRSEP) Safe Shutdown Methodology was reviewed against the requirements of Appendix R Sections IlI.G, Ill.J, and lII.L as required by 10CFR50.48(b). NRC review and approval of the HBRSEP safe shutdown methodology is contained in a series of Safety Evaluation Reports.

For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with the transition process, NEI 00-01, Revision 1 (which has since been updated to revision 2) was one of the references used in developing the circuit analysis procedure, which is now captured in FIR-NGGC-01 01, Revision 2. Except as noted in this document, the plant's methodology meets the guidelines of NEI 00-01.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 3.2, 3.34 (NSCA)

HBRSEP LAR Rev 0 Page B-2

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NLS-84-516, Fire Protection Rule - Alternate Safe Shutdown Capability -

Sections III.G.3 and ll.L of Appendix R to 10 CFR 50 - H.B. Robinson Steam Electric Plant Unit No. 2 NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10 CFR 50, Items III.G.3 and IllL; Alternate Safe Shutdown Capability - H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1 [A, Intro] Safe This section discusses the identification of systems available and necessary to perform the required Shutdown Systems safe shutdown functions. It also provides information on the process for combining these systems and Path into safe shutdown paths. Appendix R Section IIl.G.l.a requires that the capability to achieve and Development maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section III.G.1.b requires that repairs to systems and equipment necessary to achieve and maintain cold shutdown be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking.

[Refer to hard copy of NEI 00-01 for Figure 3-1]

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns RNP-E/ELEC-1216 identifies the systems and components necessary achieve and maintan safe shutdown.

For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with the transition process, NEI 00-01, Revision 1 was one of the references used in developing the circuit analysis procedure, which is now captured in FIR-NGGC-0101, revision 1. Except as noted in this document, the plant's methodology meets the guidelines of NEI 00-01.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment (NSCA)

RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.0 Nuclear Plant HBRSEP LAR Rev 0 Page B-3

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 505 Section 2.4.2.1 Nuclear Satety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1 [B, Goals] Safe The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and Shutdown Systems components remains free of fire damage for a single fire in any single plant fire area. This goal is and Path accomplished by determining those functions important to achieve and maintain hot shutdown. Safe Development shutdown systems are selected so that the capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following:

Reactivity control Pressure control systems Inventory control systems Decay heat removal systems Process monitoring Support systems

- Electrical systems

- Cooling systems These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment. Ifthese 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.

ADPlicability Comments Applicable Alignment Alignment Basis Statement Aligns RNP-E/ELEC-1216 identifies the safe shutdown functions.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-4

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

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

- A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability

- A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.

Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment. Common power source and common enclosure concems could also affect these and must be addressed.

Apmlicability Comments Applicable Alignment Statement Alignment Basis Statement Aligns HBRSEP has considered spurious operation, common power sources, and common enclosure concerns that would cause a circuit to be considered an associated circuit.

RCS isolation valves (such as the RHR Pump suction valves) are defined as high/low pressure interface boundary valves if their spurious operation could lead to the rupture of low pressure piping or a loss of RCS inventory that exceeds the RCS makeup capability. Such interface boundary valves are subject to more stringent circuit analysis criteria, and are identified in FSSPMD by the HLP flag.

This high/low pressure interface boundary valve definition is conservative with respect to that in in Appendix C of NEI 00-01 and NFPA 805 FAQ 06-0006.

During the re-validation, the definition from the previous SSA was carried forward for conservatism.

Thus, some components are classified as high-low interfaces which do not meet the above definition since their spurious opening will not result in a rupture of downstream piping and a subsequent intersystem LOCA. Robinson may choose to remove the classification of these components as high-low interfaces at a future date.

Comments HBRSEP LAR Rev 0 Page B-5

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.4, 3.34 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1 Criteria / The following criteria and assumptions may be considered when identifying systems available and Assumptions necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.

ADplicabilitv Comments Applicable Alignment Alignment Basis Statement N/A This is generic introductory information and contains no specific requirements.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.1 [GE BWR [BWR] GE Report GE-NE-T43-00002-00-01-ROl entitled "Original Safe Shutdown Paths For The Paths] BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section III.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.

Aoolicabilitv Comments HBRSEP LAR Rev 0 Page B-6

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Not Applicable Alignment

~Algnm.. Alianment Basis Statement N/A HBRSEP is a PWR. This guidance is specific to BWRs.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.10 [Manual / Manual initiation from the main control room or emergency control stations of systems required to Automatic Initiation of achieve and maintain safe shutdown is acceptable where permitted by current regulations or Systems] approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option.

Applicability Comments Applicable Aliqnment Statement Alignment Basis Aligns Reliance on the automatic logic for safe shutdown systems is not required, but if credited needs to be appropriately evaluated as being free of fire damage. The only automatic logics evaluated at HBRSEP are the Emergency Diesel Generator (EDG) Automatic Sequencing Logic and the Fast Bus Transfer logic. These logics are incorporated in the overall SSD fault tree for HBRSEP.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.4.3 and 2.2.4.4 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could HBRSEP LAR Rev 0 Page B-7

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.11 [Multiple Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and Affected Units] maintain safe shutdown for each affected unit must be demonstrated.

Applicability Comments Not Applicable Allanment Statement Alionment Basis Statement N/A Robinson is a single unit site.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.2 [SRVs / LP [BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group Systems] (BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of 10CFR50 Appendix R Sections II.G.1 and Ill.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000.

Aoplicability Comments Not Applicable Alignment SAtement Alianment Basis Statement N/A HBRSEP is a PWR. This guidance is specific to BWRs.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components HBRSEP LAR Rev 0 Page B-8

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.3 [Pressurizer [PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be Heaters] maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.

Annlicabilift Comments Applicable Alignment Alignment Basis Aligns with intent In most fire areas, HBRSEP does not rely on the use of pressurizer heaters to maintain hot shutdown.

RCS pressure is controlled during hot shutdown and cooldown by controlling the rate of charging to the RCS. Pressurizer heaters and/or auxiliary spray reduces operator burden. Neither component is required to provide adequate pressure control if charging is available. 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.

Use of the SI Pumps in lieu of the Charging Pumps may be required for certain shutdown scenarios.

The RCS would have to be depressurized to less than operating pressure of the Sl pumps.

Pressurizer heaters are credited to stabilize pressure transients when SI pumps are operated intermittently.

The NEI guidance does not prevent the use of pressurizer heaters, but only serves to note that they are generally not required.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.1.2 (2)

Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could HBRSEP LAR Rev 0 Page B-9

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 Re NEI 00-01 Guidance 3.1.1.4 [Alternative The classification of shutdown capability as alternative shutdown is made independent of the Shutdown Capability] selection of systems used for shutdown. Alternative shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections II.G.1 and III.G.2 may be supplemented by the use of manual actions to the extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations, as appropriate. These may also be used in conjunction with alternative shutdown capability.

Applicabilitl Comments Applicable Alianment Alignment Basis Statement Aligns The plant's alternate and dedicated safe shutdown systems and strategies were reviewed and approved in the supplemental SER.

Comments Reference Document Doc Detail NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10 CFR 50, Items III.G.3 and IllL; Alternate Safe Shutdown Capability - H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106 Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 R NEI 00-01 Guidance 3.1.1.5 [Initial At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains)

Conditions] are assumed operable and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress.

The units are assumed to be operating at full power under normal conditions and normal lineups.

Aoplicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns HBRSEP LAR Rev 0 Page B-10

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (2)

(NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.6 [Other Events No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, in Conjunction with earthquake), single failures or non-fire induced transients need be considered in conjunction with the Fire] fire.

Applicability Comments Applicable Alignment Alignmen Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (6,7,8)

(NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Re NEI 00-01 Guidance HBRSEP LAR Rev 0 Page B-1 1

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 3.1.1.7 [ Offsite For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire Power] damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns For fire areas that use redundant shutdown capabilities offsite power is credited unless the fire impacts equipment required to support offsite power. If the fire impacts offsite power, at least one onsite power source is available to provide the required power.

For areas that use alternative / dedicated shutdown, a LOOP is assumed.

In the analysis the LOOP is not credited for preventing or terminating spurious operations or positioning SSE in its required position. Steps in the procedures insure that the appropriate actions are taken to line up SSE and deal with potential spurious equipment operations.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.8 [Safety- Post-fire safe shutdown systems and components are not required to be safety-related.

Related Equipment]

Ainlicabillitv Comments Applicable Alignment Alignment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.

HBRSEP LAR Rev 0 Page B-12

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.1.9 [72 Hour The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor Coping] scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can be repaired or made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using onsite capability to achieve cold shutdown.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns NFPA 805 does not require a plant to transition to cold shutdown in the event of a fire. The fire area-by-fire area assessment documents the method of accomplishment of the NFPA 805 performance goals, including an optional transition to cold shutdown. For all fires at HBRSEP, the systems and equipment required to place the plant in a safe and stable condition are available following a fire occurring while the plant is at power without regard to a specific mission time or event coping duration.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 1.5.1 and 1.5.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to HBRSEP LAR Rev 0 Page B-1 3

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2 Shutdown The following discussion on each of these shutdown functions provides guidance for selecting the Functions systems and equipment required.for safe shutdown. For additional information on BWR system selection, refer to GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths for the BWR."

Applicability Comments Applicable Allonment Alignment Basis Statement Aligns This is an introductory section with no specific requirements. The GE information does not apply to HBRSEP.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.1 Reactivity [BWR] Control Rod Drive System Control The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.

[PWR] Makeup/Charging There must be a method for ensuring that adequate shutdown margin is maintained by ensuring borated water is utilized for RCS makeup/charging.

Agwlicability Comments Applicable Aliinment Alignment Basis Statemen Aligns Reactivity control is provided by insertion of control rods via a reactor trip. Long term reactivity control is provided by boron addition via the charging pumps or safety injection pumps taking suction from the Reactor Water Storage Tank.

Comments HBRSEP LAR Rev 0 Page B-14

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.2.2.5 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.2 Pressure The systems discussed in this section are examples of systems that can be used for pressure Control Systems control. This does not restrict the use of other systems for this purpose.

[BWR] Safety Relief Valves (SRVs)

The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are operated manually. Automatic initiation of the Automatic Depressurization System is not a required function.

[PWR] Makeup/Charging RCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiating charging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is acceptable.

Aoolicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns The Reactor Coolant Pressure Control function uses the same components as the RCS Inventory Control function. RCS Pressure is controlled by controlling the rate of charging to the RCS.

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. Pressurizer heaters are credited for pressure control when the SI pumps are used for makeup. When the SI pumps are utilized for makeup, the RCS is de-pressurized to the SI Pump operating pressure by cycling a Pressurizer PORV.

Comments Reference Document Doc Detail RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.2.1.2 HBRSEP LAR Rev 0 Page B-1 5

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.3 Inventory [BWR] Systems selected for the inventory control function should be capable of supplying sufficient Control reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.

[PWR]: Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown. Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.

ADolicabilitv Comments Applicable Alianment Alignment Basis Statement Aligns The RCS Inventory Control function is required to restore and maintain RCS integrity and reactor coolant makeup capability to compensate for RCS fluid losses (i.e. RCP seal leak-off) and shrinkage during cooldown. Reactor Coolant Inventory Control is accomplished by the following actions:

- RCS Isolation - RV Head Vents and Pressurizer PORVs (RCS)

- Normal Letdown Isolation (CVCS)

- Excess Letdown Isolation (CVCS)

- RHR Isolation (RHR)

- Charging/makeup from the RWST via the charging pumps or SI pumps. The SI pumps require the CCW system for cooling. (CVCS, SI, CCW, SW)

- Use of the SI Pumps for RCS Makeup requires RCS depressurization via the Pressurizer PORVs.

- RCP Seal Cooling via Seal Injection and/or Thermal Barrier Cooling (CVCS, CCW, SW)

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.2.1.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-16

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.4 Decay Heat [BWR] Systems selected for the decay heat removal function(s) should be capable of:

Removal

- Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure.

- Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool).

- Removing sufficient decay heat from the reactor to achieve cold shutdown.

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

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

- Removing sufficient decay heat from the reactor to reach cold shutdown conditions.

This does not restrict the use of other systems.

Applicability Comments Applicable Alianment SAtegment Alignment Basis Statement Aligns HBRSEP uses the Auxiliary Feedwater System (AFW) and the Main Steam System (MS) to remove decay heat from the reactor through the steam generators for Hot Shutdown. The Residual Heat Removal (RHR) System is available at temperatures below 350F and pressures less than 375 psig to remove decay heat and continue the reactor cooldown to cold shutdown.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.1.3 and 2.2.1.4 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to HBRSEP LAR Rev 0 Page B-1 7

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.5 Process The process monitoring function is provided for all safe shutdown paths. IN84-09, Attachment 1, Monitoring Section IX"Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessary to perform and control the functions specified inAppendix R Section IIIL.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN84-09 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).

Ingeneral, 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.

Apolicability Comments Applicable Alignment Alignment Basis Statement Aligns The process monitoring function is capable of providing direct reading of those process variables necessary for plant operators to perform and/or control identified safe shutdown functions. Plant monitoring instrumentation, in the context of post-fire safe shutdown operation, consists of those minimal instrument channels or local gauges/indicators necessary to monitor the operation of primary shutdown components and systems, and the operation of those components or systems that provide required support functions. The parameters to be monitored during post-fire shutdown operations, along with the credited instruments, are summarized on Table 2-2 of RNP-E/ELEC-1216.

Comments Reference Document Doc Detail RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.1.8, 2.2.2.12, and Table 2-2 HBRSEP LAR Rev 0 Page B-18

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6 Support [Blank Heading - No specific guidance]

Systems AgPlicabilitY Comments Not Applicable Algnmenn Alignment Basis St~aternent N/A Support system requirements will be addressed under the corresponding NEI 00-01 sub-section.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 806 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 806 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.1 Electrical AC Distribution System Systems Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E busses either directly from the busses or through step down transformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sources or the emergency diesel generator depending on 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 HBRSEP LAR Rev 0 Page B-19

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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).

Apelicabilit Comments Applicable Alignment Alignment Basis Statement Aligns The Electrical Distribution System provides 4160VAC, 480VAC, 120VAC and 125VDC power from off-site (115KV Grid) and onsite sources (EDGs and DSDG) to safe shutdown electrical loads. The fuel oil systems associated with the onsite power supplies are also included in the analysis.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.1.7, 2.2.2.11 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 Cooling HVAC Systems Systems [HVAC]

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).

HBRSEP LAR Rev 0 Page B-20

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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.

Applicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns Plant ventilation systems are required functional to provide environmental conditions that support continuous occupancy or safe shutdown equipment operation in the following area:

- EDG Rooms

- Main Control Room

- Motor Driven AFW Pump Room

- SI Pump Room The Main Control Room ventilation system includes refrigerant units that are cooled by the SW system.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.1.9, 2.2.2.13, Table 2-3 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result inthe maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.2.6.2 Cooling Various cooling water systems may be required to support safe shutdown system operation, based Systems [Main on plant-specific considerations. Typical uses include:

Section] - RHR/SDC/DH Heat Exchanger cooling water

- Safe shutdown pump cooling (seal coolers, oil coolers)

- Diesel generator cooling

- HVAC system cooling water HBRSEP LAR Rev 0 Page B-21

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review AoDlicabilitv Comments Applicable Aliqnment Alignment Basis Statement Aligns The Component Cooling Water and Service Water Systems provide cooling to the various safe shutdown loads.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.1.5, 2.2.1.6 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.3 Methodology for Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown Shutdown System systems and developing the shutdown paths.

Selection The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-2]

ADDlicabilitv Comments Applicable Alignment Alianment Basis statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection HBRSEP LAR Rev 0 Page B-22

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.1 Identify safe Review available documentation to obtain an understanding of the available plant systems and the shutdown functions functions required to achieve and maintain safe shutdown. Documents such as the following may be reviewed:

- Operating Procedures (Normal, Emergency, Abnormal)

- System descriptions

- Fire Hazard Analysis

- Single-line electrical diagrams

-Piping and Instrumentation Diagrams (P&IDs)

[BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR" Applicability Comments Applicable Alignment Alignment Basis Statement Aligns Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire Hazard Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system comprising the safe shutdown paths, the mechanical or electrical equipment required for the operation of the system and the equipment whose spurious operation could affect the performance of the safe shutdown systems were identified.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.1 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2.6 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

HBRSEP LAR Rev 0 Page B-23

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.2 Identify Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of Combinations of systems capable of achieving the safe shutdown functions of reactivity control, pressure control, Systems that Satisfy inventory control, decay heat removal, process monitoring, and support systems such as electrical Each Safe Shutdown and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other Function systems. In addition to achieving the required safe shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function.

ADDlicability Comments Applicable Alignment Alignment Basis Statement Aligns In accordance with the provisions of 10CFR50, Appendix R, Section IhlG, at least one means of achieving and maintaining safe shutdown conditions must remain available in the event of a fire in any fire area. In developing an appropriate shutdown equipment complement to support this requirement, it is necessary to categorize equipment into logical train-oriented groupings, identified as shutdown categories; these categories are further defined as Alternate A and Alternate B.

Although, in many cases, the equipment complement selected corresponds closely to safety-related train divisions, it should not be construed that Alternate A and B equipment automatically corresponds to safety-related Train A and B equipment. In general, the Alternate A division constitutes the equipment credited for safe shutdown outside the control room (i.e. dedicated shutdown) and the Alternate B division constitutes the equipment credited for control room shutdown scenarios.

Comments Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those crtical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.3 Define Select combinations of systems with the capability of performing all of the required safe shutdown Combinations of functions and designate this set of systems as a safe shutdown path. In many cases, safe shutdown Systems for Each paths may be defined on a divisional basis since the availability of electrical power and other support Safe Shutdown Path systems must be demonstrated for each path.

Aoolicabilitv Comments Applicable HBRSEP LAR Rev 0 Page B-24

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Alianment Alignment Basis Statement Aligns Components have been grouped into Appendix R fire-safe shutdown systems according to the Appendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is divided into one or more "success paths". Each success path represents a functionally independent method of accomplishing a unique fire-safe shutdown function. Each success path is divided into two or more redundant Appendix R fire-safe shutdown trains or "success path trains". Each success path train will often be comprised of components from different plant systems necessary to accomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safe shutdown fault tree developed during the safe shutdown re-validation project.

Comments Reference Document Doc Detail RNP-E/ELEC-1 216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.1.2.1 - 2.1.2.4, 2.2.4.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.1.3.4 Assign Assign a path designation to each combination of systems. The path will serve to document the Shutdown Paths to combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to Each Combination of this document (NEI 00-01) for an example of a table illustrating how to document the various Systems combinations of systems for selected shutdown paths.

Applicability Comments Applicable Alianment Alignment Basis Statement Aligns Components have been grouped into Appendix R fire-safe shutdown systems according to the Appendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is divided into one or more "success paths". Each success path represents a functionally independent method of accomplishing a unique fire-safe shutdown function. Each success path is divided into two or more redundant Appendix R fire-safe shutdown trains or "success path trains". Each success path train will often be comprised of components from different plant systems necessary to accomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safe shutdown fault tree developed during the safe shutdown re-validation project.

Comments HBRSEP LAR Rev 0 Page B-25

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Reference Document Doc Detail RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.2.1 - 2.2.2.4, Table 2-1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 806 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2 Safe Shutdown The previous section described the methodology for selecting the systems and paths necessary to Equipment Selection achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix R function. The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the maloperation of the safe shutdown systems.

Applicability Comments Applicable Alignment Statement Alignment Basis N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1 Criteria / Consider the following criteria and assumptions when identifying equipment necessary to perform the HBRSEP LAR Rev 0 Page B-26

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Assumptions required safe shutdown functions:

Applicability Comments Applicable Alianment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.1 [Primary 3.2.1.1 Safe shutdown equipment can be divided into two categories. Equipment may be categorized Secondary as (1) primary components or (2) secondary components. Typically, the following types of Components] equipment are considered to be primary components:

- Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.

- All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)

- Power supplies or other electrical components that support operation of primary components (i.e.,

diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).

Secondary components are typically items found within the circuitry for a primary component. These provide a supporting role to the overall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either an 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 Comments Applicable HBRSEP LAR Rev 0 Page B-27

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Alignment Alignment Basis Statement Aligns Components are not identified as primary or secondary. Components providing a "secondary" function are either identified as safe shutdown components and included in the safe shutdown equipment list, or have their applicable cables assigned to the primary component.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.2 [Fire Damage 3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact to Mechanical their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping Components (not materials, including tubing with brazed or soldered joints, are not included inthis assumption). Fire electrically damage should be evaluated with respect to the ability to manually open or close the valve should supervised)] this be necessary as a part of the post-fire safe shutdown scenario.

ARelicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns Due to the substantial nature of equipment and the nature and location of combustibles, fire will not not impact the pressure boundary function. A fire does not cause a manual valve to change its position. Manual stroking of a valve once the fire is extinguished is evaluated as part of the manual action feasibility study.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.1.3, 9.4.1 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-28

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.3 [Manual Valve Assume that manual valves are in their normal position as shown on P&IDs or in the plant operating Positions] procedures.

ARDlicabilitv Comments Applicable Alignment Statement Alianment Basis Statement Aligns This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.4 [Check Assume that a check valve closes in the direction of potential flow diversion and seats properly with Valves] sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions.

Applicabilitv Comments Applicable Alignment Alignment Basis Statement HBRSEP LAR Rev 0 Page B-29

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Aligns FIR-NGGC-001 identifies that properly oriented check valves credited as system boundaries should be included in the SSEL, and that those in the flow path need not be included. Check Valves credited as boundaries are included in the SSEL, but the assumption that they are leak tight is inherent in the analysis.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.3 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.5 [Instrument Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow Failures] transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit.

Applicability Comments Applicable Alignment Basis Statement Aligns Instruments exposed to the fire are assumed to fail. It is a generally accepted practice (that can be verified based on a review of the fire area analysis) that instruments are assumed to fail to their worst case position.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.2 and 9.4.1 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire HBRSEP LAR Rev 0 Page B-30

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.6 [Spurious Identify equipment that could spuriously operate or mal-operate and impact the performance of Components] equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.

Aoolicability Comments Applicable ASittment Alignment Basis Aligns FIR-NGGC-0101 states, "Electrically operated or controlled valves or dampers in the flow paths whose spurious operation could adversely affect system operation shall be included on the SSEL."

This is affirmed in the Section 2.1.2.5 of RNP-E/ELEC-1216.

RIS 2004-03 was a reference for the procedures used in the safe shutdown re-validation, and is also referenced in FIR-NGGC-0101.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 2.1, 9.1.3 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.1.2.5, 3.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.1.7 (instrument Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a Tubing] result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.

Apolicabilitv Comments Applicable HBRSEP LAR Rev 0 Page B-31

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Alignment Alignment Basis Statement Aligns FIR-NGGC-01 01 provides direction for evaluating the fire effects on instrument tubing and the potential impact on spurious operation. FSSPMD documents tubing routing to ensure the impact of this issue is evaluated.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.1.7 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2.7 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2 Methodology for Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown Equipment Selection equipment.

Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-3]

Agolicability Comments Applicable Alianment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection HBRSEP LAR Rev 0 Page B-32

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.1 Identify the Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each System Flow Path for shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept.

Each Shutdown Path Aoolicabllitv Comments Applicable Alignment Alignment Basis Statement Aligns The safe shutdown flow paths at Robinson are depicted on the HBR2-11390 Series of drawings.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.1.3 (NSCA)

HBR2-11390, Appendix R and Station Blackout Safe-Shutdown Analysis Flowpath/Boundary Diagrams Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.2 Identify the Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to Equipment in Each assure that all equipment in each system's flow path has been identified. Assure that any equipment Safe Shutdown that could spuriously operate and adversely affect the desired system function(s) is also identified. If System Flow Path additional systems are identified which are necessary for the operation of the safe shutdown system Including Equipment under review, include these as systems required for safe shutdown. Designate these new systems That May Spuriously with the same safe shutdown path as the primary safe shutdown system under review (Refer to Operate and Affect Figure 3-1).

System Operation Applicability Comments Applicable HBRSEP LAR Rev 0 Page B-33

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Alignment Alignment Basis Statement Aligns Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire Hazard Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system comprsing the safe shutdown paths, the mechanical or electrical equipment required for the operation of the system and the equipment whose spudous operation could affect the performance of the safe shutdown systems were identified.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.1.3 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 2.1.2.5 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.3 Develop a List Prepare a table listing the equipment identified for each system and the shutdown path that it of Safe Shutdown supports. Identify any valves or other equipment that could spuriously operate and impact the Equipment and operation of that safe shutdown system. Assign the safe shutdown path for the affected system to Assign the this equipment. During the cable selection phase, identify additional equipment required to support Corresponding the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this System and Safe additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an Shutdown Path(s) example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe Designation to Each. shutdown and it documents various equipment-related attributes used in the analysis.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns System and component identification is discussed in section 2.2.2, and section 2.2.3 refers to the SSEL maintained in FSSPMD. Attachment 24 of is a printout of the SSEL. This includes valves and pumps whose spurious operation may impact a safe shutdown system from performing its function.

Information in FSSPMD includes the component's power supply, fire zone location, normal and required positions, required cables, and associated circuits.

The components and the safe shutdown function(s) they support are depicted in the safe shutdown HBRSEP LAR Rev 0 Page B-34

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review fault tree. Development of the fault tree is described in FIR-NGGC-0101, revision 2.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.2 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.2.2, 2.2.3, and Att. 24 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.4 Identify Collect additional equipment-related information necessary for performing the post-fire safe Equipment shutdown analysis for the equipment. In order to facilitate the analysis, tabulate this data for each Information Required piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a for the Safe SSEL. Examples of related equipment data should include the equipment type, equipment Shutdown Analysis description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as the following may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, and spurious operation concem.

Aoolicabilitv Comments Applicable Alignment Aliqnment Basis Statement Aligns The information identified as needed for performing safe shutdown analysis on the components identified on the SSEL is contained in the FSSPMD. This can be verified on a component basis through reports that can be generated through the FSSPMD.

Information in FSSPMD includes the component's power supply, fire zone location, normal and required positions, required cables, and associated circuits.

Comments Reference Document Doc Detail FSSPMD, Fire Safe Shutdown Program Manager Database HBRSEP LAR Rev 0 Page B-35

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Attachment 24 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire Requirement event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Guidance 3.2.2.5 Identify In the process of defining equipment and cables for safe shutdown, identify additional supporting Dependencies equipment such as electrical power and interlocked equipment. As an aid in assessing identified Between Equipment, impacts to safe shutdown, consider modeling the dependency between equipment within each safe Supporting shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram Equipment, Safe (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these Shutdown Systems relationships.

and Safe Shutdown Paths.

Applicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns Power supplies are identified and documented in the FSSPMD. Cables that are associated with a component because of interlocks or permissive are documented with the component in the FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components HBRSEP LAR Rev 0 Page B-36

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3 Safe Shutdown This section provides industry guidance on the recommended methodology and criteria for selecting Cable Selection and safe shutdown cables and determining their potential impact on equipment required for achieving Location and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the maloperation of safe shutdown 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.

ADolicability Comments Applicable Alianment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

HBRSEP LAR Rev 0 Page B-37

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review (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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1 Criteria / To identify an impact to safe shutdown equipment based on cable routing, the equipment must have Assumptions cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment.

Consider the following criteria when selecting cables that impact safe shutdown equipment:

Applicability Comments Applicable Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend HBRSEP LAR Rev 0 Page B-38

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NE100-01 Ref NEI 00-01 Guidance 3.3.1.1 [Cable The list of cables whose failure could impact the operation of a piece of safe shutdown equipment Selection] includes more than those cables connected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, and equipment status indication cables related to the equipment. Consider reviewing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post-fire safe shutdown. As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4.

Applicability Comments Applicable Alignment Statement Alignment Basis Aligns FIR-NGGC-0101 Section 9.3.1 provides direction for assigning cables to components. This process is documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circ~its required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This HBRSEP LAR Rev 0 Page B-39

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.2 [Cables In cases where the failure (including spurious actuations) of a single cable could impact more than Affecting Multiple one piece of safe shutdown equipment, include the cable with each piece of safe shutdown Components] equipment.

Applicability Comments Applicable Alignment SAtegment Alignment Basis Statement Aligns Circuit analysis is performed independently on individual components, so cables affecting more than one component will be identified with each applicable component.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1 216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

HBRSEP LAR Rev 0 Page B-40

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review (b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.3 [Isolation Electrical devices such as relays, switches and signal resistor units are considered to be acceptable Devices] isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function.

ARRlicabilitv Comments Applicable Alignment SAtegment Alignment Basis Statement Aligns Isolation devices are defined in FIR-NGGC-0101, Section 3.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 3, Item 43.

(NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 1.3.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend HBRSEP LAR Rev 0 Page B-41

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.4 [Identify "Not Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e.,

Required" Cables] annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these circuits is necessary. However, they must be isolated from the component's control scheme in such a way that a cable fault would not impact the performance of the circuit.

APmlicability Comments Applicable Alignment Alignment Basis Statement Aligns In FSSPMD cables that are not required for safe shutdown have an "A" or an "NA" entered in the FMEA section of the circuit information form in FSSPMD. The "A" indicates that the component "achieves" its safe shutdown function even if that cable is damaged by fire. The "NA" indicates that the cable is not part of a safe shutdown circuit.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 and Attachment 1 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

HBRSEP LAR Rev 0 Page B-42

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.5 [Identification For each circuit requiring power to perform its safe shutdown function, identify the cable supplying of Power Supplies] power to each safe shutdown and/or required interlock component. Initially, identify only the power cables from the immediate upstream power source for these interlocked circuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns The power cables for individual components are listed in the circuit analysis for that component if power is needed for the component to perform its safe shutdown function. Power supplies are linked to their components in FSSPMD in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" portion of the Circuit Information Form. A standard note "A" entered for a power supply in this section indicates that the power supply is required for the component to perform its safe shutdown function. The power supply requirement is modeled in the fault tree.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 and Attachment 1 (NSCA)

RNP-E/ELEC-1 216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly HBRSEP LAR Rev 0 Page B-43

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.6 [ESFAS The automatic initiation logics for the credited post-fire safe shutdown systems are not required to Initiation] support safe shutdown. Each system can be controlled manually by operator actuation in the main control room or emergency control station. If operator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function.

ADnlicability Comments Applicable Alignment Basis Statement Aligns Reliance on the automatic logic for safe shutdown systems is not credited at HBRSEP. Although automatic ESFAS signals are not credited, they have been included in the SSA to assure that an actuation of the logic does not cause any adverse consequences.

The Fast Bus Transfer scheme and EDG Auto Sequencing fault trees have been modeled so that they may be credited when available.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection NuclearSafety Capability Assessment Sections 9.2.3, 9.3.1, 9.3.2, and 9.3.7 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.1.3.2 and 2.2.4.4 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

HBRSEP LAR Rev 0 Page B-44

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review (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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.1.7 [Circuit Cabling for the electrical distribution system is a concern for those breakers that feed associated Coordination] circuits and are not fully coordinated with upstream breakers. With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component. For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also support the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source and not the connected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.

ADolicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns The guidelines that will be used in the evaluation of the common power supplies are as follows (Ref.

FIR-NGGC-0101):

- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for electrical coordination will be developed.

- For each safe shutdown power supply, the existing short circuit calculations, load studies, coordination calculations, protective device setting sheets, and time current curves as appropriate to confirm proper coordination between upstream and downstream protective devices will be reviewed.

- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be considered.

- For cases in which coordination between series protective devices cannot be demonstrated, a common power supply associated circuit will be assumed to exist. These circuits will be dispositioned by one of the following means:

1) Demonstrate coordination by refining the available short circuit current and/or trip device characteristics.

2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,

equipment located in same fire area as power supply).

3) Identify readily achievable protective device setting changes (including changes in fuse size and/or clearing characteristics) that will establish coordination.

4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 10CFR50 Appendix R safe shutdown when protection devices do not provide the desired coordination.

5) Existing short circuit and coordination calculations will be updated as necessary to fully document HBRSEP LAR Rev 0 Page B-45

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review where coordination is credited for 10CFR50 Appendix R safe shutdown.

The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault on the same breaker's load cable would prevent the breaker from tripping on over-current and could result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV power supply circuit analysis as Associated Circuits and Cables.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA)

FPP-RNP-200, 10CFR50, Appendix R,Section 111.G, Associated Circuits Analysis RNP-E-8.005, 10CFR50 Appendix RAssociated Circuit, Common Power Supply Analysis RNP-E-8.053, Non-Safety Overcurrent Protection Coordination RNP-E-9.021, 10CFR50 Appendix RFuse Analysis for DS Bus RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.2.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance HBRSEP LAR Rev 0 Page B-46

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

- Spurious actuations

- Common power source

- Common enclosure.

Aoolicabilitv Comments Applicable Alianment Alilnment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [A] Associated Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose Circuit Cables - failure could cause the spurious actuation/mal-operation of equipment whose operation could affect Cables Whose safe shutdown. These cables are identified in Section 3.3.3 together with the remaining safe Failure May Cause shutdown cables required to support control and operation of the equipment.

HBRSEP LAR Rev 0 Page B-47

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Spurious Actuations Apolicabillitv Comments Applicable Alinnmenl Alignment Basis Statement Aligns Cables that can cause an undesired spurious actuation are identified by an "S"in the FMEA code of the circuit information form in FSSPMD. They are evaluated in the SSA in the same manner as "required" cables. RNP-E/ELEC-1216 evaluates throughout for spurious operation of valves, pumps, and breakers.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Sections 3.0, 9.1.3, 9.3.2, and Attachment 1 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.2.3 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits, that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [B] Associated The concern for the common power source associated circuits is the loss of a safe shutdown power Circuit Cables - source due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on a Common Power HBRSEP LAR Rev 0 Page B-48

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Source Cables 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.

Applicabilitv Comments Applicable Alianment Alignment Basis Statement Aligns The guidelines that will be used in the evaluation of the common power supplies are as follows (Ref.

FIR-NGGC-0101):

- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for electrical coordination will be developed.

- For each safe shutdown power supply, the existing short circuit calculations, load studies, coordination calculations, protective device setting sheets, and time current curves as appropriate to confirm proper coordination between upstream and downstream protective devices will be reviewed.

- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be considered.

- For cases in which coordination between series protective devices cannot be demonstrated, a common power supply associated circuit will be assumed to exist. These circuits will be dispositioned by one of the following means:

1) Demonstrate coordination by refining the available short circuit current and/or trip device characteristics.

2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,

equipment located in same fire area as power supply).

3) Identify readily achievable protective device setting changes (including changes in fuse size and/or clearing characteristics) that will establish coordination.

4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 10CFR50 Appendix R safe shutdown when protection devices do not provide the desired coordination.

5) Existing short circuit and coordination calculations will be updated as necessary to fully document where coordination is credited for 10CFR50 Appendix R safe shutdown.

The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault on the same breaker's load cable would prevent the breaker from tripping on over-current and could result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV power supply circuit analysis as Associated Circuits and Cables.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA)

RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common Power HBRSEP LAR Rev 0 Page B-49

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Supply Analysis RNP-E-8.053, Non-Safety Overcurrent Protection Coordination RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.2.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.2 [C] Associated The concern with common enclosure associated circuits is fire damage to a cable whose failure Circuit Cables - could propagate to other safe shutdown cables in the same enclosure either because the circuit is Common Enclosure not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result Cables in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area.

This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e.,

multiple fires). Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.

ApplIcability Comments Applicable Alignment Alignment Basis Statement Aligns The following guidelines were used in the evaluation of common enclosure associated circuits (Ref.

FIR-NGGC-0101):

- Perform an evaluation of the common enclosure associated circuits by reviewing design and HBRSEP LAR Rev 0 Page B-50

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review installation criteria for cable and electrical penetrations. Confirm that cables are adequately protected against short circuits and will not propagate a fire from one fire area to another. In evaluating common power supply circuits the acceptance criteria shall not be limited to standard cable damage temperatures, which are based on not degrading cable insulation (typically 250 0C for thermoset cable). Rather, the criteria will be based on not exceeding temperatures at which self ignition or damage to surrounding cables could occur.

- If a common enclosure associated circuit is determined to exist, the concern shall be resolved by one of the following means:

1) Demonstrate by analysis that the cable does not pose a risk to cables within the common enclosure under fault conditions (i.e., the cable exceeds its recommended temperature rise but does not represent a hazard to surrounding cables),

2) Demonstrate that the lack of fault protection does not adversely affect safe shutdown,

3) Identify readily achievable protective device setting changes (including changes in fuse size and/or time-current characteristics) that will establish cable protection without affecting other performance requirements, or

4) Incorporate the cables of concern into the safe shutdown analysis as post-fire safe shutdown cables for the affected power supply.

5) Existing short circuit and electrical protection calculations will be updated as necessary.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA)

FPP-RNP-200, 10CFR50, Appendix R, Section IlI.G, Associated Circuits Section 4.0 Analysis RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.2.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly HBRSEP LAR Rev 0 Page B-51

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 Re NEI 00-01 Guidance 3.3.3 Methodology for Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables Cable Selection and necessary for performing a post-fire safe shutdown analysis.

Location Use the following methodology to define the cables required for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-4]

Aoolicability Comments Applicable Alianment Statement Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

HBRSEP LAR Rev 0 Page B-52

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.1 Identify For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical Circuits Required for diagrams including the following documentation to identify the circuits (power, control, the Operation of the instrumentation) required for operation or whose failure may impact the operation of each piece of Safe Shutdown equipment:

Equipment - 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.

Ifpower 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.

ADDlicabilitv Comments Applicable Alianment Statement Alignment Basis Aligns FIR-NGGC-0101 Section 9.3.2 provides direction for assigning cables to components. This process is further documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss of HBRSEP LAR Rev 0 Page B-53

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.2 Identify In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, Interlocked Circuits cables and equipment. Assign to the equipment any cables for interlocked circuits that can affect the and Cables Whose equipment.

Spurious Operation or While investigating the interlocked circuits, additional equipment or power sources may be Mal-operation Could discovered. Include these interlocked equipment or power sources in the safe shutdown equipment Affect Shutdown list (refer to Figure 3-3) if they can impact the operation of the equipment under consideration.

Aoolicabilitv Comments Applicable Alianment Alignmen Alignment Basis Statement Aligns with intent As an alternative to adding the interlocked equipment to the SSEL, it is acceptable to include the cables that are required for the interlocking function (or that could cause the spurious actuation) with the main component that was originally under consideration. Adding them to the components may ease the development of a suitable mitigating strategy in areas where the interlocked cables may be damaged by the fire. Interlocked circuits were either included in the analysis, or the interlocked contact or relay was assumed to be in its worst-case position. Associated circuits identified for each component are either included in the main circuit analysis, or are included by listing the applicable circuit in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" on the Circuit Information Form.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

HBRSEP LAR Rev 0 Page B-54

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.3 Assign Cables Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that to the Safe Shutdown may result in maloperation of each piece of safe shutdown equipment.

Equipment Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, the same cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component.

If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged.

Apolicability Comments Applicable Alionment Alignment Basis Statement Aligns The circuit analysis results for each electrically operated safe shutdown component are contained in the FSSPMD. FSSPMD contains various forms and reports for presenting SSD cables and associated circuits. Refer to Progress Energy procedure FIR-NGGC-0101 for a description of the circuit analysis nomenclature used in the FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.2 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.2.1.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.

NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified.

HBRSEP LAR Rev 0 Page B-55

Attachment 8- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.4 Identify Identify the routing for each cable including all raceway and cable endpoints. Typically, this Routing of Cables information is obtained from joining the list of safe shutdown cables with an existing cable and raceway database Applicability Comments Applicable Aliqnment tatement Alignment Basis Statement Aligns Equipment location and Cable raceway routing information (i.e. cable raceway to fire area/zone correlation) was migrated in 2004 from the safe shutdown analysis of record at the time (FPP-RNP-150, Revision 7A) to the FSSPMD. Additional cables were added to FSSPMD based on revised component selection.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.9 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 4.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.

NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified.

NEI 00-01 Ref NEI 00-01 Guidance 3.3.3.5 Identify Identify the fire area location of each raceway and cable endpoint identified in the previous step and Location of Raceway join this information with the cable routing data. In addition, identify the location of field-routed cable and Cables by Fire by fire area. This produces a database containing all of the cables requiring fire area analysis, their Area locations by fire area, and their raceway.

Apolicabilitv Comments Applicable Alianment Basis Statement Aligns Cable to raceway information is contained in the Cable Information Forms of the FSSPMD.

Raceway and endpoint locations for all required cables are also contained in FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.9 HBRSEP LAR Rev 0 Page B-56

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 4.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 806 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4 Fire Area By determining the location of each component and cable by fire area and using the cable to Assessment and equipment relationships described above, the affected safe shutdown equipment in each fire area Compliance can be determined. Using the list of affected equipment in each fire area, the impacts to safe Assessment shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined.

The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document.

Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individually mitigating the effects of each of the potential impacts.

AoRlicabilitv Comments Applicable Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1 Criteria / The following criteria and assumptions apply when performing fire area compliance assessment to Assumptions mitigate the consequences of the circuit failures identified in the previous sections for the required HBRSEP LAR Rev 0 Page B-57

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review safe shutdown path in each fire area.

Applicabilitv Comments Applicable Alianment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.1 [Number of Assume only one fire in any single fire area at a time.

Postulated Fires]

Aoolicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns RNP-E/ELEC-1216 postulates only one fire occurring at a time.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 1.4 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

HBRSEP LAR Rev 0 Page B-58

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.2 [Damage to Assume that the fire may affect all unprotected cables and equipment within the fire area. This Unprotected assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the Equipment and exposure fire that is required by the regulation.

Cables]

Applicability Comments Applicable Aliqnment Statement Alignment Basis Aligns RNP-E/ELEC-1216 postulates all electrical equipment and cables in a given fire area are damaged and unavailable unless NRC exemption or appropriate evaluation (GL-86-10) has been completed.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 5.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.3 [Assess Address all cable and equipment impacts affecting the required safe shutdown path in the fire area.

Impacts to Required All potential impacts within the fire area must be addressed. The focus of this section is to determine Components] and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected.

Amplicabilitv Comments Applicable Alignment Alignment Basis Statement Aligns All potential impacts of the fire are identified in the fault tree. Potential damage to equipment required to show success in each area is addressed with an appropriate compliance strategy. The results are documented in FSSPMD and in RNP-E/ELEC-1216.

Comments Reference Document Doc Detail HBRSEP LAR Rev 0 Page B-59

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 5.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.4 [Manual Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in Actions] accordance with NRC requirements.

Applicabilitv Comments Applicable Alianment Alignment Basis Statement Aligns Manual actions credited in the shutdown analysis are summarized on a fire area basis in Attachment 26 of RNP-E/ELEC-1216. RNP-E-8.050 documents the feasibility of the manual actions. The current regulatory guidance, as reflected in FAQs 06-0012 and 07-0030 was used as the basis for determining the acceptability of the manual actions.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Attachment 2 (NSCA)

RNP-E-8.050, Appendix R Transient Analysis and Timeline Evaluation for H.B. Robinson - Unit No. 2 RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 5.2 and Att. 26 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance HBRSEP LAR Rev 0 Page B-60

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 3.4.1.5 [Repairs] Where appropriate to achieve and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, use repairs to equipment required in support of post fire shutdown.

Applicability Comments Applicable Allignmen Alignment Basis Statemet Aligns Repairs are identified where necessary for cold shutdown equipment. A list of credited repair procedures can be found in RNP-E/ELEC-1216.

No repairs are required to achieve safe and stable conditions.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.15 and 9.4.2 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 1.7.2.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.6 [Assess Appendix R compliance requires that one train of systems necessary to achieve and maintain hot Compliance with shutdown conditions from either the control room or emergency control station(s) is free of fire Deterministic Criteria] damage (lll.G.l.a). When cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s):

- Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (lll.G.2.a)

- Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (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 non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (llI.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 HBRSEP LAR Rev 0 Page B-61

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review noncombustible radiant energy shield (lII.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.

Applicabilitv Comments Applicable Alianment Alignment Basis Statement Aligns This section of NEI 00-01 repeats the requirements of Appendix R III.G.2. RNP-E/ELEC-1216 documents how each fire area has adequate systems to comply with these requirements and the requirements of NFPA 805 Sections 4.2.3 and 4.2.4 for the post-transition configuration.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.2 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 5.0 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.7 [Consider Consider selecting other equipment that can perform the same safe shutdown function as the Additional Equipment] impacted equipment. In addressing this situation, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPP's current licensing basis.

Applicability Comments Applicable Alianment Alignment Basis Statement Aligns This consideration is not clearly stated but is inherent in performing a safe shutdown analysis. RNP-E/ELEC-1216 only documents the systems and components that were actually selected and not those that were considered but not necessary.

Any plant system that supports meeting the safe shutdown performance goals may be considered for inclusion in the SSEL. However, the intent is to minimize the systems and components identified in the SSEL for configuration control purposes. If the system or component cannot directly assist in demonstrating compliance with the deterministic requirements of NFPA 805, its inclusion in the SSEL may not be warranted.

HBRSEP LAR Rev 0 Page B-62

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.1 and 9.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.1.8 [Consider Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent Instrument Tubing effects on instrument readings or signals associated with the protected safe shutdown path in Effects] evaluating post-fire safe shutdown capability. This can be done systematically or via procedures such as Emergency Operating Procedures.

Apolicabilitv Comments Applicable Alignment Statement Aliqnment Basis Aligns RNP-E/ELEC-1216 documents the consideration of fire effects on instrument tubing for HBRSEP.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.1.7 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Sections 2.1.2.7, 2.2.2.12, Att. 22, and Att. 25 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

HBRSEP LAR Rev 0 Page B-63

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.4.2 Methodology for Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area Fire Area assessment.

Assessment Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance:

[Refer to hard copy of NEI 00-01 for Figure 3-5]

Applicability Comments Applicable Alignment Alianment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.1 Identify the Identify the safe shutdown cables, equipment and systems located in each fire area that may be Affected Equipment potentially damaged by the fire. Provide this information in a report format. The report may be by Fire Area sorted by fire area and by system in order to understand the impact to each safe shutdown path within each fire area (see Attachment 5 for an example of an Affected Equipment Report).

Aoolicabllitv Comments Applicable Aligqnment Alianment Basis Statement Aligns RNP-E/ELEC-1216 lists the components potentially affected in each fire area. These reports are available in the FSSPMD.

Comments Reference Document Doc Detail FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1 216, The Fire Safe Shutdown Analysis for H.B. Robinson Attachment 22 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-64

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.2 Determine the Based on a review of the systems, equipment and cables within each fire area, determine which Shutdown Paths shutdown paths are either unaffected or least impacted by a postulated fire within the fire area.

Least Impacted By a Typically, the safe shutdown path with the least number of cables and equipment in the fire area Fire in Each Fire Area would be selected as the required safe shutdown path. Consider the circuit failure criteria and the possible mitigating strategies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of this assessment since their availability will be important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require this power to support their operation.

Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious operation or mal-operation could affect the shutdown function. Include these cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path.

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.

ADolicabllitv Comments Applicable Alignment Alignment Basis Statement Aligns RNP-E/ELEC-1216 identifies both the equipment that is selected for a given fire area and the equipment that is not selected. This shows that the division selected for a given safe shutdown system is the train that was generally that least affected by the fire.

The use of a fault tree in the analysis helps to ensure that components that may have support systems affected, such as cooling or power supplies, are not credited without taking these failures into account.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4 HBRSEP LAR Rev 0 Page B-65

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Att. 22 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.3 Determine Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine Safe Shutdown the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire Equipment Impacts area, and what those possible impacts are.

Applicability Comments Applicable Alionment Alignment Basis Statement Aligns RNP-E/ELEC-1216 identifies the equipment used for safe shutdown and what the potential impact of the fire on the safe shutdown equipment. This information is also contained in FSSPMD.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Attachment 22 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.4 Develop a The available deterministic methods for mitigating the effects of circuit failures are summarized as Compliance Strategy HBRSEP LAR Rev 0 Page B-66

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review or Disposition to follows (see Figure 1-2):

Mitigate the Effects - Provide a qualified 3-fire rated barrier.

Due to Fire Damage - Provide a 1-hour fire rated barrier with automatic suppression and detection.

to Each Required - Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate Component or Cable 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.

ADDiicabilitv Comments Applicable Alignmen Alignment Basis Statement Aligns RNP-E/ELEC-1216 verifies that appropriate separation is used for redundant cables. This can take the form of 3-hour fire barriers, 1-hour fire barriers with suppression and detection, or 20 feet of separation with suppression and detection. In some cases, exemptions have been requested from and granted by the NRC for configurations that did not meet these requirements. Also some fire protection evaluations have determined that the protection in place provides adequate separation for the hazards of the area.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Attachment 22 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement . requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.4.2.5 Document the Assign compliance strategy statements or codes to components or cables to identify the justification Compliance Strategy or mitigating actions proposed for achieving safe shutdown. The justification should address the or Disposition cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide Determined to each piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-Mitigate the Effects operation could affect safe shutdown, and/or cable for the required safe shutdown path with a HBRSEP LAR Rev 0 Page B-67

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Due to Fire Damage specific compliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area to Each Required Assessment Report documenting each cable disposition.

Component or Cable Applicabilitv Comments Applicable Alianment tiatement Alignment Basis Statement Aligns The fire area by fire area separation reports contained in Attachment 22 of RNP-E/ELEC-1216 identify all the analyzed circuits and components and the credited compliance strategies.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Attachment 22 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5 Circuit Analysis This section on circuit analysis provides information on the potential impact of fire on circuits used to and Evaluation monitor, control and power safe shutdown equipment. Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with HBRSEP LAR Rev 0 Page B-68

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 IIl.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits.

ARolicabilitv Comments Applicable AlInmen Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required by NFPA 805 Sections 4.2.3 and 4.2.4 for the post-transition configuration.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1 Criteria Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.

Assumptions Aoplicability Comments Applicable Alignment Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific HBRSEP LAR Rev 0 Page B-69

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.1 [Circuit Failure Consider the following circuit failure types on each conductor of each unprotected safe shutdown Types and Impact] cable to determine the potential impact of a fire on the safe shutdown equipment associated with that conductor.

- A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor. A hot short may cause a spurious operation of safe shutdown equipment.

- An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit may prevent the ability to control or power the affected equipment. An open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs]

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

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

Consider the three types of circuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area.

HBRSEP LAR Rev 0 Page B-70

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Applicabiley Comments Applicable Alignment Alignment Basis Statement Aligns The safe shutdown circuit analysis shall be reviewed and updated as necessary for credible circuit failures as a deterministic analysis utilizing the Current Design Method (CDM). These failures include:

- Multiple shorts to ground or grounded conductor.

- Multiple open circuits.

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

- Cable-to-cable shorts are postulated to occur.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.2 [Circuit Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal Contacts and mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst HBRSEP LAR Rev 0 Page B-71

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Operational Modes] must consider the position of the safe shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns Components are assumed to be in their normal position at the time of the fire. This includes electrical contacts and switches.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 9.3.2 and 9.4.1 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.3 [Duration of Assume that circuit failure types resulting in spurious operations exist until action has been taken to Circuit Failures] isolate the given circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indicates that fire-induced hot shorts typically self-mitigate after a limited period of time.

HBRSEP LAR Rev 0 Page B-72

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review ApplicabilitY Comments Applicable Alianment tatemet Alignment Basis Statement Aligns "Hot Short" duration is considered to exist until action has been taken to isolate the given circuit from the fire area, or other actions as appropriate have been taken to negate the effects of the spurious actuation. HBRSEP does not postulate that the fire will eventually clear the "Hot Short."

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3.6 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NE100-01 Ref NEI 00-01 Guidance 3.5.1.4 [Cable Failure When both trains are in the same fire area outside of primary containment, all cables that do not Configurations] meet the separation requirements of Section III.G.2 are assumed to fail in their worst case configuration.

Applicability Comments Applicable Alignment Alignment Basis Statement HBRSEP LAR Rev 0 Page B-73

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Aligns The following damage is assumed to occur due to the postulated fire:

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

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

Electrical equipment located in a fire area is assumed to fail as a result of the postulated fire in the fire area, and is considered unavailable to ensure completion of safe shutdown functions unless it meets the separation criteria of 10 CFR 50 Appendix R or is shown to be acceptable as-is based on an approved exemption. This electrical equipment includes motors, instruments, I/P converters, controllers, switches, MCC's, switchgear, transformers, generators, batteries, panel boards, etc.

Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.4.1 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 0"I0 Ref NEI 00-01 Guidance 3.5.1.5 [A, Circuit The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to Failure Risk identify any potential combinations of spurious operations with higher risk significance. Bin 1 failures Assessment should also be the focus of the analysis; however, NRC has indicated that other types of failures Guidance] required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be followed.

HBRSEP LAR Rev 0 Page B-74

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Applicability Comments Not Applicable Alignment Alignment Basis Statement NIA This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5 [B, Cable For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the Failure Modes] same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It 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 HBRSEP LAR Rev 0 Page B-75

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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. Iffire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.

Applicability Comments Not Applicable Aliqnment Alignment Basis Statement Aligns Section 9.3.3 of FIR-NGGC-0101 describes the specific cable failure modes to be considered in conducting a circuit analysis. Section 3.0 of FIR-NGGC-0101 contains general definitions for the cable failure modes. Configurations considered included the following, and were applied to all safe shutdown cables.

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

b) Inter-cable shorting of thermoset cables, or thermoset and thermoplastic cables, are considered to be credible events.

c) Compatible polarity hot shorts for DC circuits were considered to the degree specified in the cases below:

- Case 1 - Intra-Cable Shorts within a Single Cable For this case, a single cable must contain both a source and target conductor for both polarities. It is postulated that intra-cable shorts within the cable will result in compatible polarity connections for both polarities (e.g., a plus-to-plus and a minus-to-minus connection for a DC control circuit). Given the relatively high probability of intra-cable conductor-to-conductor shorting, this failure mode was considered.

- Case 2 - Intra-Cable Shorts on Separate Cables For this case, two independent but coincident hot shorts of the proper polarity (without grounding) in separate cables must occur. Given the relative high probability of intra-cable conductor-to-conductor HBRSEP LAR Rev 0 Page B-76

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review shorting, this failure mode was considered.

- Case 3 - Inter-Cable Shorts on Separate Cables For this case two independent but coincident hot shorts of the proper polarity (without grounding) must occur. This case differs from Case 1 and 2 in that one or both of the hot shorts must involve inter-cable shorting. Given the low likelihood of coincident proper polarity shorts combined with the low likelihood of inter-cable hot shorting, this failure mode was only considered for components identified as "high-low pressure interface" or Fire PRA "high consequence equipment."

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

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

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

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

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.0, 9.3.2, 9.3.3, 9.3.10, 9.4.3, 9.4.6 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 1.1.2.3 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the Requirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref NEI 00-01 Guidance 3.5.1.5 [C, Likelihood Determination of the potential consequence of the damaged associated circuits is based on the of Undesired examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components Consequences] that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown.

When considering the potential consequence of such failures, the [analyst] should also consider the time at which the prevented operation or maloperation occurs. Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation.

Consideration of cold-shutdown circuits is deferred pending additional research.

Applicability Comments Applicable Alianment Statement Alignment Basis HBRSEP LAR Rev 0 Page B-77

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Aligns with intent RNP-E/ELEC-1216 limits the evaluation of multiple spurious operations, implementing the design strategy of any and all potential spurious operation, on a one at a time basis. As part of the manual action feasibility study, two concurrent spurious operations were evaluated.

Multiple spurious operations (MSO) were considered for a variety of scenarios by the MSO Expert panel. Components were identified for consideration and possible inclusion in the Safe Shutdown Analysis and the Fire PRA. Any MSOs that were determined to be risk-significant by the PRA were analyzed accordingly.

Comments Reference Document Doc Detail RNP-E-8.050, Appendix R Transient Analysis and Timeline Evaluation for H.B. Robinson - Unit No. 2 RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2 Types of Circuit Appendix R requires that nuclear power plants must be designed to prevent exposure fires from Failures defeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of HBRSEP LAR Rev 0 Page B-78

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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.

Apolicability Comments Applicable Alignment Statement Alignment Basis Statement N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

Comments Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NE100-01 Ref NEI 00-01 Guidance 3.5.2.1 Circuit This section provides guidance for addressing the effects of an open circuit for safe shutdown Failures Due to an equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit Open Circuit continuity. An open circuit will typically prevent the ability to control or power the affected HBRSEP LAR Rev 0 Page B-79

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 Comments Applicable Alignment Basis Statemient Aligns Circuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method (RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes multiple open circuits.

The potential CT circuits of concern have been identified, and the final disposition of this potential fire scenario will be assessed as part of the SSA/Fire PRA transition to NFPA 805. Refer to Implementation item in Attachment "S", Table S-3.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3, 9.3.4 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.1.4 Nuclear Plant HBRSEP LAR Rev 0 Page B-80

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit This section provides guidance for addressing the effects of a short-to-ground on circuits for safe Failures Due to a shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system Short-to-Ground [A, resulting in the potential on the conductor being applied to ground potential. A short-to-ground can General] cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist.

Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:

- A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment.

- In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.

Aoolicabilitv Comments Applicable Alignment Alignment Basis Statemen N/A This is an introductory statement and provides no requirements. The sub-paragraphs with specific requirements are addressed separately as required.

HBRSEP LAR Rev 0 Page B-81

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.2 Circuit This section provides guidance for addressing the effects of a short-to-ground on circuits for safe Failures Due to a shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system Short-to-Ground [B, resulting in the potential on the conductor being applied to ground potential. A short-to-ground can Grounded Circuits] cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist.

Short-to-Ground on Grounded Circuits Typically, in the case of a grounded circuit, a short-to-ground on any part of the circuit would present a concern for tripping the circuit isolation device thereby causing a loss of control power.

Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-2]

Short-to-ground No. 1:

A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit. This will result an inability to operate the equipment using the control switch.

Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.

HBRSEP LAR Rev 0 Page B-82

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review Short-to-ground No. 2:

A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.

Applicability Comments Applicable Alignment Alignment Basis Statement Aligns Circuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method (RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes multiple shorts to ground.

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.1.4 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance HBRSEP LAR Rev 0 Page B-83

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 3.5.2.2 Circuit Short-to-Ground on Ungrounded Circuits Failures Due to a Short-to-Ground [C, In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the Ungrounded Circuits] circuit may not result in tripping the circuit isolation device. Another short-to-ground on the circuit or another circuit from the same source would need to exist to cause a loss of control power to the circuit.

Figure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-3]

Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the same circuit or on any other circuit fed from the same power source. This will result in an inability to operate the equipment using the control switch. Depending on the coordination characteristics between the protective device on this circuit and upstream circuits, the power supply to other circuits could be affected.

Short-to-ground No. 2:

A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch is closed. Should this occur, the effect would be identical to that for the short-to-ground at location No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop control switch, the equipment will still be able to be opened/started.

Applicability Cmet Applicable Allgnment Allanment Basis Statement Aligns A single ground fault on an ungrounded AC or DC control circuit has no immediate functional affect.

Thus, ungrounded systems are more resilient to functional failures. Nonetheless, multiple ground faults are credible and must be considered. For ease of analysis, an existing - but unspecified -

ground fault from the same power source will be assumed when analyzing ungrounded systems.

Furthermore, multiple shorts-to-ground are to be evaluated for their impact on ungrounded circuits.

As noted in FIR-NGGC-0101, it is likely that over the course of a fire at least one conductor from each polarity of a circuit (positive and negative polarity) will eventually become grounded. Thus, the circuit analysis should not try to take credit for a circuit remaining functional simply because two conductors must short to ground to render the circuit inoperable (i.e., blow the fuse or trip the circuit breaker).

Comments Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-84

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse).is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit This section provides guidance for analyzing the effects of a hot short on circuits for required safe Failures Due to a Hot shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between Short [A, General] conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.

Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:

- A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.

- A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to-cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.

APPlicability Comments Applicable Alianment Alianment Basis Statement Aligns The Current Design Method is the safe shutdown circuit analysis method used for applying failures to circuits. One key attribute of this method is that a hot short is applied independent of the cable configuration and is applied as a hot probe. The probe's power is postulated to be present, and its HBRSEP LAR Rev 0 Page B-85

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review source is not identified.

Unless otherwise documented, all cables at HBRSEP were assumed to be thermo-plastic. Hot shorts are postulated to occur regardless of the cable insulation type.

Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Sections 3.17 and 9.3.3 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit A Hot Short on Grounded Circuits Failures Due to a Hot Short [B, Grounded A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground as Circuits] described above would result in de-energizing the circuit. This would further reduce the likelihood for the circuit to change the state of the equipment either from a control switch or due to a hot short.

Nevertheless, a hot short still needs to be considered. Figure 3.5.2-4 shows a typical grounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section. In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The following discussion describes how to address the impact of these individual cable faults on the operation of the equipment controlled by HBRSEP LAR Rev 0 Page B-86

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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.

ADDlicability Comments Applicable Alignment Alignment Basis Statement Aligns Hot shorts on grounded circuits were considered. Cables susceptible to grounds were identified with the associated equipment.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA)

Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals.

This will ensure that a comprehensive population of circuitry is evaluated.

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.

(a) Common Power Supply Circuits. Those circuits whose 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 if the upstream protection device (i.e., breaker or fuse) is not properly coordinated with the downstream protection device.

(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.

HBRSEP LAR Rev 0 Page B-87

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.3 Circuit A Hot Short on Ungrounded Circuits Failures Due to a Hot Short [C, Ungrounded In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spurious Circuits] operation. A single hot short can cause a spurious operation if the hot short comes from a circuit from the positive leg of the same ungrounded source as the affected circuit.

In reviewing each of these cases, the common denominator is that in every case, the conductor in the circuit between the control switch and the start/stop coil must be involved.

Figure 3.5.2-5 depicted below shows a typical ungrounded control circuit that might be used for a motor-operated valve. However, the protective devices and position indication lights that would normally be included in the control circuit for a motor-operated valve have been omitted, since these devices are not required to understand the concepts being explained in this section.

In the discussion provided below, it is assumed that a single fire in a given fire area could cause any one of the hot shorts depicted. The discussion provided below describes how to address the impact of these cable faults on the operation of the equipment controlled by this circuit.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-5]

Hot short No. 1:

A hot short at this location from the same control power source would energize the close relay and result in the undesired closure of a motor operated valve.

Hot short No. 2:

A hot short at this location from the same control power source would energize the open relay and result in the undesired opening of a motor operated valve.

Applicability Comments Applicable Alignment Stategment Alignment Basis Statement Aligns Hot shorts on ungrounded circuits were considered. Cables susceptible to grounds were identified with the associated equipment. Hot shorts between two cables were considered credible in the RNP analysis.

Comments Reference Document Doc Detail FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Assessment Section 9.3.3 (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.1.1 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.

HBRSEP LAR Rev 0 Page B-88

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.4 Circuit The evaluation of associated circuits of a common power source consists of verifying proper Failures Due to coordination between the supply breaker/fuse and the load breakers/fuses for power sources that are Inadequate Circuit required for safe shutdown. The concern is that, for fire damage to a single power cable, lack of Coordination coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment.

For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump.

To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump, which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due to the associated circuits' common power source breaker coordination issue, load breaker 4 must be fully coordinated with the feeder breaker to the Train A bus.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-6]

A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be considered to ensure that coordination is demonstrated at the maximum fault level.

The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:

- Identify the power sources required to supply power to safe shutdown equipment.

- For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus.

- For each power source, demonstrate proper circuit coordination using acceptable industry methods.

- For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown. Prepare a list of the following information for each fire area:

- Cables of concern.

- Affected common power source and its path.

- Raceway in which the cable is enclosed.

- Sequence of the raceway in the cable route.

- Fire zone/area in which the raceway is located.

HBRSEP LAR Rev 0 Page B-89

Attachment B- NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review 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 Comments Applicable Alignment Alignment Basis Statement Aligns Power cables for Safe Shutdown equipment have been selected for evaluation for all components that are required to change states. Coordination of electrical breakers and fuses assure that other power cables from loads on the same electrical bus or distribution center will not adversely impact safe shutdown equipment. FIR-NGGC-0101 provides guidance on verifying that circuit coordination exists, as well as methods for addressing cases where coordination is not readily apparent.

Circuit breaker and fuse coordination are verified by calculations.

CommentS Reference Document Doc Detail FIR-NGGC-01 01, Fire Protection Nuclear Safety Capability Assessment Section 9.3.6 (NSCA)

RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common Power Supply Analysis RNP-E-8.053, Non-Safety Overcurrent Protection Coordination RNP-E-9.021, 10CFR50 Appendix R Fuse Analysis for DS Bus RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.2.2 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 806 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.

NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be Requirement identified.

NEI 00-01 Ref NEI 00-01 Guidance 3.5.2.5 Circuit The common enclosure associated circuit concern deals with the possibility of causing secondary Failures Due to failures due to fire damage to a circuit either whose isolation device fails to isolate the cable fault or Common Enclosure protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates Concerns along the cable into adjoining fire areas.

The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of HBRSEP LAR Rev 0 Page B-90

Attachment B-NEI 04-02 Table B-2 Nuclear Safety Duke Energy Capability Assessment Methodology Review the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.

ApRlicablt Comments Applicable Alignmet Alianment Basis Statement Aligns Circuit breaker and fuse coordination are verified by calculations. Adequate coordination exists to assure that a common enclosure issue is not credible.

Commengs Reference Document Doc Detail FPP-RNP-200, 10CFR50, Appendix R, Section Ill.G, Associated Circuits Section 4.0 Analysis RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Robinson Section 3.2.2.3 Nuclear Plant Table B-2 Nuclear Safety Capability Assessment Methodology Review HBRSEP LAR Rev 0 Page B-91

Duke Energy Attachment E - Radioactive Release Transition Duke Energy Attachment E Radioactive Release Transition E. NEI 04-02 Radioactive Release Transition 24 Pages Attached Page E-1 HBRSEP LAR Rev 0 LAR Rev 0 Page E-1

Duke Energy Attachment E - Radioactive Release Transition Radioactive Release Analysis Compartmentation Auxiliary Building and Radwaste Building Areas Auxiliary and Radwaste Building Areas each have floor drains which are routed to various sumps providing containment and monitoring prior to release. Auxiliary Building normal and emergency ventilation discharge is monitored prior to release. Ventilation is typically secured during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003. Auxiliary Building run-off drains to the settling ponds. The settling ponds are monitored regularly by chemistry and would be tested after a fire event that potentially had a radiological release. Air samples are taken by an HP on duty during a fire event.

Containment The Containment Building has floor drains routed to various containment building sumps which provide containment and monitoring prior to release. Containment ventilation systems during modes 1-4 provide a recirculation path with discharge that is monitored prior to release. Containment and monitoring of ventilation during Modes 5, 6 and Defueled are administered by plant procedure OMM-033, Implementation of CV Closure. HBRSEP employs a protective membrane that may be deployed to secure the CV Maintenance Hatch in the event of a fire during outage periods when this barrier is open.

Fuel Handling Building (FHB) Areas Fuel Handling Building has floor drains which route to sump(s) providing containment and monitoring prior to release. FHB ventilation exhaust via HVE-14, 15, and 15A is filtered and monitored via Radiation Monitors R-14, R-20, and R-21.

Turbine Building General Building is open to the outdoors. Areas other than the CV Access were screened out for Radioactive Release potential. CV Access areas have floor drains which are routed to various sumps providing containment and monitoring prior to release.

Miscellaneous Buildings and Components Miscellaneous areas were reviewed on a case-by-case basis for drains and ventilation provisions for containment and monitoring based on the potential for the presence of radioactive materials Generic Assumptions / Discussions

• The project instruction FPIP-0121, "Radiological Release Reviews During Fire Fighting Activities" was utilized.

" A review of Access / Egress pathway guidance for potential smoke release paths was conducted and a review of ventilation and smoke removal and recommended actions included within Table E-1.

" Under the generic Fire Pre-Plan outline the Hazards Section identifies typical fixed radiological hazards for each area.

• Engineering controls such as use of forced air ventilation and damming for fire suppression agent run off was considered for each fire pre-plan and recommendations are included within Table E-1.

" Sea-Land type containers are used for temporary storage of radioactive materials outside of the normal RCA area and these are considered outside the scope of this review due to their location.

" The applicable limits of 10 CFR 20, Appendix B Table 2 Column 1 were considered.

" Radioactive materials in transport were not considered.

HBRSEP LAR Rev 0 Page E-2

Duke Energy Attachment E - Radioactive Release Transition Training Review HBRSEP has transitioned its fire brigade training lesson plans to NFPA 600 compliant lessons. This review was conducted to identify and validate the adequacy of attributes associated with the fleet standard NFPA 600 lesson plans to address the Radioactive Release Performance Criteria associated with NFPA 805. Fire Brigade lesson plans which provide training guidance and information relative to Radioactive Release are identified below; 1- FPI0001G-08, Rev 0, Safety and Orientation

• Slide #10 introduces NFPA 805 and its performance criteria applicable to the fire brigade including the radiological release criteria.

2- FPI0001G-02, Rev 0, Personnel Protective Equipment

• Slide #22 addresses the process for cleaning of tools and gear used by the fire brigade and off site responders to ensure no radioactively contaminated material is released after a fire event.

3- FPI0001G-10, Rev 0, Fire Hose

• Slide #49 provides instruction for the containment and monitoring for potentially radioactively contaminated fire suppression agents.

4- FPI000IG-03, Rev 0, Forcible Entry

• Slide #5 provides instruction for the containment and monitoring for potentially radioactively contaminated fire combustion products.

5- FPI0001G-12, Rev 0, Ventilation

• Slide #20 provides instruction for the containment and monitoring for potentially radioactively contaminated fire combustion products.

10- FPI0001G-04, Rev 0, Salvage & Overhaul

• Slide #10 provides instruction for the containment and monitoring for potentially radioactively contaminated fire suppression agents and products of combustion.

12- FPI000IG-07, Rev 0, Fire Attack

• Slide #38 provides instruction for the containment and monitoring for potentially radioactively contaminated fire suppression agents and products of combustion for areas such as hose line placement.

No additional changes were required to the fire brigade training materials Page E-3 HBRSEPLARRevO HBRSEP LAR Rev 0 Page E-3

Duke Energy Attachment E - Radioactive Release Transition Pre-Fire Plan Review Pre-fire plans were reviewed to determine which features are in place to prevent or minimize an uncontrolled radiological release due to a fire event or firefighting activities.

Specifically, this review included a description of the radiological hazards, the drainage and water containment features present, HVAC systems present, and the potential for cross-contamination of radiologically clean areas due to fire fighting activities and fire suppression agents such as water, foam and portable fire extinguishers (CO 2 , dry chemical, etc.).

Engineered Controls Review Drainage information was derived from drainage design basis documentation. The location of floor drains were reviewed to determine if drain paths lead to proper filtering and monitoring of liquid radioactive waste before release, consistent with regulatory limits. HVAC and radiation monitoring design basis documentation was reviewed to determine which areas featured HVAC systems designed to contain and process airborne contamination. Pre-fire plan and station fire protection plan drawings were reviewed to determine which areas have the potential for cross-contamination of a radiological boundary due to firefighting activities.

The results of the radioactive review are documented in the following Table E-1.

Page E-4 HBRSEPLARRevO HBRSEP LAR Rev 0 Page E-4

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Diesel Generator Al/ 8.1, 8.10 Screened In Auxiliary Building Areas have floor FZ "B" EDG, OMM- The performance "B" Room FZ-1 drains which are routed to various 003 Section 8.10 requirements of sumps providing containment and In Step 8.10.3.6.c add note NFPA 805 of monitoring prior to release. to tie back to Section 5.5 Radiological release Auxiliary Building normal and for controls of water runoff. will be satisfied with the emergency ventilation discharge is revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

ESR 94-00952 modified the floor drains to the outside. Drawing HBR2-07452 shows this drain to the outside as grouted closed.

Diesel Generator A2/ 8.1, 8.9 Screened In Auxiliary Building Areas have floor FZ "A" EDG. OMM- The performance "A" Room FZ-2 drains which are routed to various 003 Section 8.9 requirements of sumps providing containment and In Step 8.9.3.6.c add note NFPA 805 of monitoring prior to release. to tie back to Section 5.5 Radiological release Auxiliary Building normal and for controls of water runoff. will be satisfied with the emergency ventilation discharge is revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

ESR 94-00952 modified the floor drains to the outside. Drawing HBR2-07452 shows this drain to the outside as grouted closed.

HBRSEP LAR Rev 0 Page E-5

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Areal Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Safety Injection A9/ 8.2, 8.16 Screened In Auxiliary Building Areas have floor FZ SI Pump Room, The performance Pump Room FZ-3 drains which are routed to various OMM-003 Section 8.16 requirements of sumps providing containment and Investigate reversing the NFPA 805 of monitoring prior to release. Primary Attack Routes - Radiological release Auxiliary Building normal and need to look at SSA will be satisfied with the emergency ventilation discharge is Shutdown Requirements. revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured In Step 8.16.3.6.b add note Training Materials.

during fire events as directed by to tie back to Section 5.5 AOP-041. On-scene containment for controls of water runoff.

and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003 Charging Pump B/ 8.1, 8.6 Screened In Auxiliary Building Areas have floor FZ Charging Pump The performance Room, VCT FZ-4 drains which are routed to various Room, OMM-003 Section requirements of NFPA Room sumps providing containment and 8.6 805 of Radiological monitoring prior to release. In Step 8.6.3.6.c add note release will be satisfied Auxiliary Building normal and to tie back to Section 5.5 with the revision of Pre-emergency ventilation discharge is for controls of water runoff. Fire Plans and existing monitored prior to release. Training Materials Ventilation is typically secured during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003 Page E-6 LAR Rev HBRSEP LAR HBRSEP Rev 00 Page E-6

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Component C/ 8.1, 8.7 Screened In Auxiliary Building Areas have floor FZ CCW Pump Room., The performance Cooling Pump FZ-5 drains which are routed to various OMM-003 Section 8.7 requirements of Room sumps providing containment and Investigate reversing the NFPA 805 of monitoring prior to release. Primary Attack Routes - Radiological release Auxiliary Building normal and need to look at SSA will be satisfied with the emergency ventilation discharge is Shutdown Requirements. revision of Pre-Fire monitored prior to release. In Step 8.7.3.6.c add note Plans and existing Ventilation is typically secured to tie back to Section 5.5 Training Materials.

during fire events as directed by for controls of water runoff.

AOP-041. On-scene containment Suggest breaking this and/or monitoring of smoke and bullet into sub-components water run-off in the fire affected - one specifically for water area is directed by the Fire runoff, one for chromates, Protection Pre-plan, OMM-003. another for oil, etc.

Auxiliary A6/ 8.43 Screened Out Miscellaneous areas were FZ Auxiliary Not Required Feedwater Pump FZ-6 reviewed on a case-by-case basis Feedwater Pump Room, Room for drains and ventilation OMM-003 Section 8.43 provisions for containment and Outside RCA. Screened monitoring based on the potential for the presence of radioactive out.

materials.

Auxiliary Bldg. A3/ 8.1, 8.2, 8.3 Screened In Auxiliary Building Areas have floor FZ-7 -Aux. Bldg. Hallway The performance Hallway (Ground FZ-7 drains which are routed to various (Ground Floor), OMM-003 requirements of Floor) sumps providing containment and Section 8.1. 8.2 and 8.3 NFPA 805 of monitoring prior to release. In Step 8.1.3.6.c add note Radiological release Auxiliary Building normal and to tie back to Section 5.5 will be satisfied with the emergency ventilation discharge is for controls of water runoff. revision of Pre-Fire monitored prior to release. In Step 8.2.3.6.e add note Plans and existing Ventilation is typically secured to tie back to Section 5.5 Training Materials.

during fire events as directed by for controls of water runoff.

AOP-041. On-scene containment In Step 8.3.3.6.c add note and/or monitoring of smoke and to tie back to Section 5.5 water run-off in the fire affected for controls of water runoff.

area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-7 HBRSEPLARRevO HBRSEP LAR Rev 0 Page E-7

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Boron Injection A8/ 8.32 Screened In Auxiliary Building Areas have floor FZ B.I.T. Room, OMM- The performance Tank Room FZ-8 drains which are routed to various 003 Section 8.32 requirements of sumps providing containment and In Step 8.32.3.6.c add note NFPA 805 of monitoring prior to release. to tie back to Section 5.5 Radiological release Auxiliary Building normal and for controls of water runoff. will be satisfied with the emergency ventilation discharge is revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

North Cable Vault D/ 8.4 Screened In Auxiliary Building Areas have floor FZ North Cable Vault, The performance FZ-9 drains which are routed to various OMM-003 Section 8.4 requirements of sumps providing containment and In Step 8.4.3.6.c add note NFPA 805 of monitoring prior to release. to tie back to Section 5.5 Radiological release Auxiliary Building normal and for controls of water runoff. will be satisfied with the emergency ventilation discharge is revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-8 LAR Rev HBRSEP LAR HBRSEP Rev 00 Page E-8

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section South Cable E/ 8.5 Screened In Auxiliary Building Areas have floor FZ South Cable The performance Vault FZ-10 drains which are routed to various Vault, OMM-003 Section requirements of sumps providing containment and 8.5 NFPA 805 of monitoring prior to release. In Step 8.5.3.6.c add note Radiological release Auxiliary Building normal and to tie back to Section 5.5 will be satisfied with the emergency ventilation discharge is for controls of water runoff. revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Pipe Alley All/ 8.12 Screened In Auxiliary Building Areas have floor FZ Pipe Alley, OMM- The performance FZ-11 drains which are routed to various 003 Section 8.12 requirements of sumps providing containment and In Step 8.12.3.6.c add note NFPA 805 of monitoring prior to release. to tie back to Section 5.5 Radiological release Auxiliary Building normal and for controls of water runoff. will be satisfied with the emergency ventilation discharge is revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-9 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page E-9

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Waste Holdup A12/ 8.15 Screened In Auxiliary Building Areas have floor FZ Waste Holdup The performance Tank, RHR Heat FZ-12 drains which are routed to various Tank RHR Heat requirements of Exchangers sumps providing containment and Exchangers, OMM-003 NFPA 805 of monitoring prior to release. Section 8.15 Radiological release Auxiliary Building normal and In Step 8.15.3.6.c add note will be satisfied with the emergency ventilation discharge is to tie back to Section 5.5 revision of Pre-Fire monitored prior to release. for controls of water runoff. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Chemical Storage A7/ 8.22 Screened In Auxiliary Building Areas have floor FZ Chemical Storage The performance Area/ Boric Acid FZ-1 3 drains which are routed to various Area/ Boric Acid Bathing requirements of Batching Tank sumps providing containment and Tank, OMM-003 Section NFPA 805 of monitoring prior to release. 8.22 Radiological release Auxiliary Building normal and Add Step 8.15.3.6.c with will be satisfied with the emergency ventilation discharge is standard statement revision of Pre-Fire monitored prior to release. concerning water runoff Plans and existing Ventilation is typically secured and note to tie back to Training Materials.

during fire events as directed by Section 5.5 for controls of AOP-041. On-scene containment water runoff.

and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-1O HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page E-10

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Solid Waste A4/ 8.28 Screened In Auxiliary Building Areas have floor FZ Solid Waste The performance Handling Room FZ-14 drains which are routed to various Handling Room, OMM- requirements of sumps providing containment and 003 Section 8.28 NFPA 805 of monitoring prior to release. In Step 8.28.3.6.d add note Radiological release Auxiliary Building normal and to tie back to Section 5.5 will be satisfied with the emergency ventilation discharge is for controls of water runoff. revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Aux. Bldg Second A5/ 8.18 Screened In Auxiliary Building Areas have floor FZ Aux. Bldg Second The performance Level FZ-15 drains which are routed to various Level, OMM-003 Section requirements of sumps providing containment and 8.18 NFPA 805 of monitoring prior to release. In Step 8.18.3.6.d add note Radiological release Auxiliary Building normal and to tie back to Section 5.5 will be satisfied with the emergency ventilation discharge is for controls of water runoff. revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Battery Room A13/ 8.20 Screened Out Miscellaneous areas were FZ Battery Room, Not Required FZ-16 reviewed on a case-by-case basis OMM-003 Section 8.20 for drains and ventilation Screened out because not provisions for containment and in Radiation Control Area.

monitoring based on the potential Evaluate excluding for the presence of radioactive monitoring for airborne materials, contamination currently required per procedure.

HBRSEP LAR Rev 0 Page E-11I

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section HVAC Equipment A14/ 8.26 Screened Out Miscellaneous areas were FZ HVAC Equipment Not Required Room for Control FZ-17 reviewed on a case-by-case basis Room for Control Room, Room for drains and ventilation OMM-003 Section 8.26 provisions for containment and Outside RCA. Screened monitoring based on the potential out.

for the presence of radioactive materials.

Unit 1 Cable A10/ 8.23 Screened Out Miscellaneous areas were FZ Unit 1 Cable Not Required Spreading Room FZ-18 reviewed on a case-by-case basis Spreading Room, OMM-for drains and ventilation 003 Section 8.23 provisions for containment and In Step 8.23.3.5.c remove monitoring based on the potential statement concerning for the presence of radioactive airborne contamination.

materials. Area is outside RCA.

Screened out.

Unit 2 Cable A15/ 8.24 Screened Out Miscellaneous areas were FZ Unit 2 Cable Not Required Spreading Room FZ-19 reviewed on a case-by-case basis Spreading Room, OMM-for drains and ventilation 003 Section 8.24 provisions for containment and In Step 8.24.3.5.c remove monitoring based on the potential statement concerning for the presence of radioactive airborne contamination.

materials. In Step 8.24.3.6.a remove statement "Low Level contamination possible."

Area is outside RCA.

Screened out.

Page E-12 HBRSEP LAR HEIRSEP Rev 0 LAR Rev 0 , Page E-1 2

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Emergency A16/ 8.25 Screened Out Auxiliary Building Areas have floor FZ Emergency Not Required Switchgear Room FZ-20 drains which are routed to various Switchgear Room and and Electrical sumps providing containment and Electrical Equipment Equipment Room monitoring prior to release. Room, OMM-003 Section Auxiliary Building normal and 8.25 emergency ventilation discharge is Area is outside RCA.

monitored prior to release. Screened out.

Ventilation is typically secured Evaluate excluding during fire events as directed by monitoring for airborne AOP-041. On-scene containment contamination currently and/or monitoring of smoke and required per procedure.

water run-off in the fire affected Clarify in 8.25.3.5.b to area is directed by the Fire monitor for airborne Protection Pre-plan, OMM-003. contamination with fires on the RCA side only.

Add Step 8.25.3.6.c with standard statement concerning water runoff and note to tie back to Section 5.5 for controls of water runoff.

Rod Control A17/ 8.27 Screened In Auxiliary Building Areas have floor FZ Rod Control The performance Room FZ-21 drains which are routed to various Room, OMM-003 Section requirements of sumps providing containment and 8.27 NFPA 805 of monitoring prior to release. Add Step 8.27.3.6.c with Radiological release Auxiliary Building normal and standard statement will be satisfied with the emergency ventilation discharge is concerning water runoff revision of Pre-Fire monitored prior to release. and note to tie back to Plans and existing Ventilation is typically secured Section 5.5 for controls of Training Materials.

during fire events as directed by water runoff.

AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-13 HBRSEP LAR HBRSEP Rev 00 LAR Rev Page E-13

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Control Room A18/ 8.30 Screened Out Miscellaneous areas were FZ-22-Control Room, Not Required FZ-22 reviewed on a case-by-case basis OMM-003 Section 8.30 for drains and ventilation Area is outside RCA.

provisions for containment and Screened out.

monitoring based on the potential for the presence of radioactive materials.

Hagan Room A18/ 8.31 Screened Out Miscellaneous areas were FZ-23-Hagan Room, Not Required FZ-23 reviewed on a case-by-case basis OMM-003 Section 8.31 for drains and ventilation Area is outside RCA.

provisions for containment and Screened out.

monitoring based on the potential Remove statement for the presence of radioactive concerning monitoring of materials, airborne contamination from Step 8.31.3.5.b.

Containment F/ 8.29 Screened In The Containment Building has FZ Containment, The performance FZ-24 floor drains routed to various OMM-003 Section 8.29 requirements of containment building sumps which Add Step 8.29.3.6.e to NFPA 805 of provide containment and consider reestablishing Radiological release monitoring prior to release. containment integrity thru will be satisfied with the Containment ventilation systems use of OMM-033, revision of Pre-Fire during modes 1-4 provide a Implementation of CV Plans and existing recirculation path with discharge Closure. Training Materials.

that is monitored prior to release.

Containment and monitoring of ventilation during Modes 5, 6 and Defueled are administered by plant procedure OMM-033, Implementation of CV Closure.

HBRSEP employs a protective membrane that may be deployed to secure the CV Maintenance Hatch in the event of a fire during outage periods when this barrier is open.

Page E-14 HBRSEP LAR HBRSEP Rev 00 LAR Rev Page E-14

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Turbine Bldg. G1/ 8.11, 8.42, 8.49 Screened In General Building is open to the FZ-25A- Turbine Building The performance East, DS FZ-25A outdoors. Areas other than the CV Ground Floor, East Side, requirements of Transformer, SW Access were screened out for OMM-003 Section 8.42, NFPA 805 of Pit #3, SW Pipe Radioactive Release potential. CV 8.49 Radiological release Enclosure Access areas have floor drains Turbine Building is outside will be satisfied with the which are routed to various sumps the RCA and thus revision of Pre-Fire providing containment and Screened Out. However, Plans and existing monitoring prior to release. part of the area within Fire Training Materials.

Zone 25A is inside the RCA (part of Old RCA Access Area and Inside AO's Office) and thus Screened In.

Turbine Bldg. GI/ 8.42, 8.44, 8.48 Screened In General Building is open to the FZ-25B- Turbine Building The performance West, CV Access FZ-25B outdoors. Areas other than the CV Ground Floor, West Side, requirements of Building Access were screened out for OMM-003 Section 8.42, NFPA 805 of Radioactive Release potential. CV 8.44, 8.48 and 8.8 Radiological release Access areas have floor drains Expand Section 8.8 to will be satisfied with the which are routed to various sumps include the RCA Entrance revision of Pre-Fire providing containment and Area (RCA Entrance Plans and existing monitoring prior to release. Tunnel Area) with Training Materials.

precautions for Radiological Hazards, water runoff statement, and airborne contamination sampling. This area would be Screened In.

The rest of the Turbine Building Ground Floor-West Side, "A"& "B" Aux.

Boilers and Associated Fuel Oil Pumps and the Makeup Water Treatment Room are outside the RCA and thus Screened Out.

Page E-15 HBRSEP Rev 0 LAR Rev HIBRSEP LAR 0 Page E-16

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM.003 Section Turbine Bldg. CV G1/ 8.8, 8.14 Screened In General Building is open to the FZ-25C- Turbine Bldg. The performance Access and RCA FZ-25C (RCA Areas) outdoors. Areas other than the CV CV Access and RCA Tool requirements of Tool Room Access were screened out for Room, OMM-003 Section NFPA 805 of Radioactive Release potential. CV 8.8. 8.14 Radiological release Access areas have floor drains In Step 8.8.3.6.c add note will be satisfied with the which are routed to various sumps to tie back to Section 5.5 revision of Pre-Fire providing containment and for controls of water runoff. Plans and existing monitoring prior to release. In Step 8.14.3.6.c add note Training Materials.

to tie back to Section 5.5 for controls of water runoff.

Dedicated G7/ 8.46 Screened Out Miscellaneous areas were FZ-25D- Dedicated Not Required Shutdown Diesel FZ-25D reviewed on a case-by-case basis Shutdown Diesel Generator for drains and ventilation Generator, OMM-003 provisions for containment and Section 8.46 monitoring based on the potential Area is outside RCA.

for the presence of radioactive Screened out.

materials.

Turbine Bldg. Gi/ 8.52, 8.53 Screened Out General Building is open to the FZ-25E- Turbine Bldg. Not Required Mezzanine Floor FZ-25E outdoors. Areas other than the CV Mezzanine Floor East and East and 4160 Access were screened out for 4160 Switchgear Room, Switchgear Room Radioactive Release potential. CV OMM-003 Section 8.52 &

Access areas have floor drains 8.53 which are routed to various sumps Area is outside RCA.

providing containment and Screened out.

monitoring prior to release.

Turbine Bldg. Gi/ 8.52 Screened Out General Building is open to the FZ-25F- Turbine Bldg. Not Required Mezzanine Floor FZ-25F outdoors. Areas other than the CV Turbine Bldg. Mezzanine West Access were screened out for Floor West, OMM-003 Radioactive Release potential. CV Section 8.52 Access areas have floor drains Area is outside RCA.

which are routed to various sumps Screened out.

providing containment and monitoring prior to release.

Page E-16 HBRSEP LAR Rev 0 LAR Rev 0 Page E-16

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Turbine Bldg. Gi/ 8.54 Screened Out General Building is open to the FZ-25G- Turbine Bldg. Not Required Operating Floor FZ-25G outdoors. Areas other than the CV ODeratinci Floor, OMM-Access were screened out for 003 Section 8.54 Radioactive Release potential. CV Area is outside RCA.

Access areas have floor drains Screened out.

which are routed to various sumps providing containment and monitoring prior to release.

Station GI/ 8.51 Screened Out Miscellaneous areas were FZ Station Not Required Transformers FZ-26 reviewed on a case-by-case basis Transformers, OMM-003 for drains and ventilation Section 8.51 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

RHR Pump Room H/ 8.38 Screened In Miscellaneous areas were FZ RHR Pit Room, The performance (RHR Pit) FZ-27 reviewed on a case-by-case basis OMM-003 Section 8.38 requirements of for drains and ventilation In Step 8.38.3.6.c add note NFPA 805 of provisions for containment and to tie back to Section 5.5 Radiological release monitoring based on the potential for controls of water runoff. will be satisfied with the for the presence of radioactive revision of Pre-Fire materials. Plans and existing Training Materials.

New and Spent G4/ 8.37, 8.39 Screened In Fuel Handling Building has floor FZ-28A- New and Spent The performance Fuel Storage FZ-28A drains which route to sump(s) Fuel Storage Areas, requirements of Areas, Heat providing containment and OMM-003 Section 8.37, NFPA 805 of Exchrs. and monitoring prior to release. FHB 8.39 Radiological release Spent Fuel Pool ventilation exhaust via HVE-14, In Step 8.37.3.6.b add note will be satisfied with the 15, and 15A is filtered and to tie back to Section 5.5 revision of Pre-Fire monitored via Radiation Monitors for controls of water runoff. Plans and existing R-14, R-20, and R-21. Add Step 8.39.3.6.c with Training Materials.

standard statement concerning water runoff and note to tie back to Section 5.5 for controls of water runoff.

HBRSEP LAR Rev 0 Page E-17

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Gas Decay Tank G4/ 8.40 Screened In Fuel Handling Building has floor FZ-28B- Gas Decay Tank The performance and Cask FZ-28B drains which route to sump(s) and Cask Preparation requirements of Preparation providing containment and Areas, OMM-003 Section NFPA 805 of Areas monitoring prior to release. FHB 8.40 Radiological release ventilation exhaust via HVE-14, In Steps 8.40.3.6.d, will be satisfied with the 15, and 15A is filtered and 8.40.3.6.e and 8.40.3.6.f revision of Pre-Fire monitored via Radiation Monitors add note to tie back to Plans and existing R-14, R-20, and R-21. Section 5.5 for controls of Training Materials.

water runoff.

CVCS Hold-Up G4/ 8.33 Screened In Fuel Handling Building has floor FZ-28C- CVCS Hold-Up The performance Tank Area FZ-28C drains which route to sump(s) Tank Area, OMM-003 requirements of providing containment and Section 8.33 NFPA 805 of monitoring prior to release. FHB In Step 8.33.3.6.d add note Radiological release ventilation exhaust via HVE-14, to tie back to Section 5.5 will be satisfied with the 15, and 15A is filtered and for controls of water runoff. revision of Pre-Fire monitored via Radiation Monitors Plans and existing R-14, R-20, and R-21. Training Materials.

Hot Machine G4/ 8.36 Screened In Fuel Handling Building has floor FZ-28D- Hot Machine The performance Shop FZ-28D drains which route to sump(s) Shop, OMM-003 Section requirements of providing containment and 8.36 NFPA 805 of monitoring prior to release. FHB Add Step 8.36.3.6.b with Radiological release ventilation exhaust via HVE-14, standard statement will be satisfied with the 15, and 15A is filtered and concerning water runoff revision of Pre-Fire monitored via Radiation Monitors and note to tie back to Plans and existing R-14, R-20, and R-21. Section 5.5 for controls of Training Materials.

water runoff.

Fuel Handling G4/ 8.34 Screened In Fuel Handling Building has floor FZ-28E- Fuel Handling The performance Building HVF-15 FZ-28E drains which route to sump(s) Building HVF-15 Fan requirements of Fan Room providing containment and Room, OMM-003 Section NFPA 805 of monitoring prior to release. FHB 8.34 Radiological release ventilation exhaust via HVE-14, Add Step 8.34.3.6.d with will be satisfied with the 15, and 15A is filtered and standard statement revision of Pre-Fire monitored via Radiation Monitors concerning water runoff Plans and existing R-14, R-20, and R-21. and note to tie back to Training Materials.

Section 5.5 for controls of water runoff.

HBRSEP LAR Rev 0 Page E-18

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Fuel Handling G4/ 8.35 Screened In Fuel Handling Building has floor FZ-28F- Fuel Handling The performance Building HVE-14 FZ-28F drains which route to sump(s) Building HVE-14 Fan requirements of Fan Room providing containment and Room, OMM-003 Section NFPA 805 of monitoring prior to release. FHB 8.35 Radiological release ventilation exhaust via HVE-14, Add Step 8.34.3.6.d with will be satisfied with the 15, and 15A is filtered and standard statement revision of Pre-Fire monitored via Radiation Monitors concerning water runoff Plans and existing R-14, R-20, and R-21. and note to tie back to Training Materials.

Section 5.5 for controls of water runoff.

Intake Structure G3/ 8.41 Screened Out Miscellaneous areas were FZ Intake Structure, Not Required FZ-29 reviewed on a case-by-case basis OMM-003 Section 8.41 for drains and ventilation Area is outside RCA.

provisions for containment and Screened out.

monitoring based on the potential for the presence of radioactive materials.

Diesel Oil G2/ 8.47 Screened Out Miscellaneous areas were FZ Diesel Oil Storage Not Required Storage Tank FZ-30 reviewed on a case-by-case basis Tank, OMM-003 Section for drains and ventilation 8.47 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Refueling Water G4/ 8.17 Screened In Miscellaneous areas were FZ Refueling Water The performance Storage Tank FZ-31 reviewed on a case-by-case basis Storage Tank, OMM-003 requirements of for drains and ventilation Section 8.17 NFPA 805 of provisions for containment and Screened in based on Radiological release monitoring based on the potential water runoff. will be satisfied with the for the presence of radioactive revision of Pre-Fire materials. Plans and existing Training Materials.

Page E-19 HBRSEP LAR HBRSEP Rev 00 LAR Rev Page E-19

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Primary Water G4/ 8.17 Screened In Miscellaneous areas were FZ Primary Water The performance Storage Tank FZ-32 reviewed on a case-by-case basis Storage Tank, OMM-003 requirements of for drains and ventilation Section 8.17 NFPA 805 of provisions for containment and Screened in based on Radiological release monitoring based on the potential water runoff. will be satisfied with the for the presence of radioactive revision of Pre-Fire materials. Plans and existing Training Materials.

Condensate GI/ 8.45 Screened Out Miscellaneous areas were FZ Condensate Not Required Storage Tank FZ-33 reviewed on a case-by-case basis Storage Tank, OMM-003 for drains and ventilation Section 8.45 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Battery "C" G4/ 8.21 Screened Out Miscellaneous areas were FZ Battery "C" Not Required Enclosure FZ-34 reviewed on a case-by-case basis Enclosure, OMM-003 for drains and ventilation Section 8.21 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Radwaste Facility G4/ 8.13 Screened In Radwaste Building has floor drains FZ Radwaste Facility, The performance FZ-35 which are routed to various sumps OMM-003 Section 8.13 requirements of providing containment and Add Step 8.13.3.6.c with NFPA 805 of monitoring prior to release. standard statement Radiological release Ventilation is typically secured concerning water runoff will be satisfied with the during fire events as directed by and note to tie back to revision of Pre-Fire AOP-041. On-scene containment Section 5.5 for controls of Plans and existing and/or monitoring of smoke and water runoff. Training Materials.

water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-20 HBRSEP LAR HBRSEP Rev 00 LAR Rev Page E-20

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section CCW Surge Tank Al 9/ 8.31 Screened Out Miscellaneous areas were FZ CCW Surcie Tank Not Required Room FZ-36 reviewed on a case-by-case basis Room, OMM-003 Section for drains and ventilation 8.31 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive Remove statement materials, concerning monitoring of airborne contamination from Step 8.31.3.5.b.

Radiation Monitor G4/ Screened In Auxiliary Building Areas have floor FZ Radiation Monitor The performance Room FZ-37 drains which are routed to various Room requirements of sumps providing containment and Incorporate into OMM-003, NFPA 805 of monitoring prior to release. Fire Pre-Plans with Radiological release Auxiliary Building normal and standard statements will be satisfied with the emergency ventilation discharge is concerning airborne revision of Pre-Fire monitored prior to release. contamination and water Plans and existing Ventilation is typically secured runoff. Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Page E-21 HBRSEP LAR Rev 00 LAR Rev Page E-21

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Waste Evaporator A4/ 8.19 Screened In Auxiliary Building Areas have floor FZ Waste Evaporator The performance Room FZ-38 drains which are routed to various Room, OMM-003 Section requirements of sumps providing containment and 8.19 NFPA 805 of monitoring prior to release. In Step 8.19.3.6.c add note Radiological release Auxiliary Building normal and to tie back to Section 5.5 will be satisfied with the emergency ventilation discharge is for controls of water runoff. revision of Pre-Fire monitored prior to release. Plans and existing Ventilation is typically secured Training Materials.

during fire events as directed by AOP-041. On-scene containment and/or monitoring of smoke and water run-off in the fire affected area is directed by the Fire Protection Pre-plan, OMM-003.

Purge Inlet Room G4/ Screened In Miscellaneous areas were FZ Purge Inlet Room The performance FZ-39 reviewed on a case-by-case basis Incorporate into OMM-003, requirements of for drains and ventilation Fire Pre-Plans with NFPA 805 of provisions for containment and standard statements Radiological release monitoring based on the potential concerning airborne will be satisfied with the for the presence of radioactive contamination and water revision of Pre-Fire materials, runoff. Plans and existing Training Materials.

Unit 2 Fuel Oil G4/ Screened Out Miscellaneous areas were FZ Unit 2 Fuel Oil Not Required Unloading/ FZ-40 reviewed on a case-by-case basis Unloading/Transfer Area Transfer Area for drains and ventilation Area is outside RCA.

provisions for containment and Screened out.

monitoring based on the potential for the presence of radioactive materials.

Deepwell Pump B YARD/ Screened Out Miscellaneous areas were FZ Deepwell Pump B Not Required area SW corner FZ-42 reviewed on a case-by-case basis area SW corner of Unit 2 of Unit 2 for drains and ventilation Switchyard Switchyard provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

HBIRSEP LAR Rev 0 Page E-22

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Unit 2 Switchyard G5/ 8.51 Screened Out Miscellaneous areas were FZ Unit 2 Switchyard, Not Required FZ-41 reviewed on a case-by-case basis OMM-003 Section 8.51 for drains and ventilation Area is outside RCA.

provisions for containment and Screened out.

monitoring based on the potential for the presence of radioactive materials.

Oil Dispensing YARD/ Screened In Miscellaneous areas were FZ Oil Dispensing The performance Building FZ-43 (RCA Areas) reviewed on a case-by-case basis Building, located in requirements of for drains and ventilation Outbuilding Pre-Plan NFPA 805 of provisions for containment and Pull the Outbuilding Pre- Radiological release monitoring based on the potential Plan into OMM-003, Fire will be satisfied with the for the presence of radioactive Pre-Plans with standard revision of Pre-Fire materials, statements concerning Plans and existing airborne contamination and Training Materials.

water runoff.

Secondary YARD/ 8.50 Screened Out Miscellaneous areas were FZ Secondary Not Required Sampling FZ-44 reviewed on a case-by-case basis Sampling, OMM-003 for drains and ventilation Section 8.50 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Northern Sheet YARD/ Screened In Miscellaneous areas were FZ Northern Sheet The performance Metal Building FZ-45 reviewed on a case-by-case basis Metal Building Adiacent requirements of Adjacent to RAB for drains and ventilation to RAB and PWST NFPA 805 of and PWST provisions for containment and Incorporate into OMM-003, Radiological release monitoring based on the potential Fire Pre-Plans with will be satisfied with the for the presence of radioactive standard statements revision of Pre-Fire materials, concerning airborne Plans and existing contamination and water Training Materials.

runoff.

Page E-23 LAR Rev HBRSEP LAR HEIRSEP Rev 00 Page E-23

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pro-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section Condensate YARD/ 8.45 Screened Out Miscellaneous areas were FZ Condensate Not Required Polisher Building FZ-46 reviewed on a case-by-case basis Polisher Building, OMM-for drains and ventilation 003 Section 8.45 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Deepwell Pump A YARD/ Screened Out Miscellaneous areas were FZ Deepwell Pump A Not Required area South of FZ-47 reviewed on a case-by-case basis area South of building building 160 for drains and ventilation 160 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Deepwell Pump C YARD/ Screened Out Miscellaneous areas were FZ Deepwell Pump C Not Required area north side of FZ-48 reviewed on a case-by-case basis area South of building building 325 for drains and ventilation 160 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

Deepwell Pump D YARD/ Screened Out Miscellaneous areas were FZ Deepwell Pump D Not Required enclosure FZ-49 reviewed on a case-by-case basis area South of building for drains and ventilation 160 provisions for containment and Area is outside RCA.

monitoring based on the potential Screened out.

for the presence of radioactive materials.

PAP East YARD/ Screened Out Miscellaneous areas were FZ PAP East Not Required FZ-50 reviewed on a case-by-case basis Area is outside RCA.

for drains and ventilation Screened out.

provisions for containment and monitoring based on the potential for the presence of radioactive materials.

HBRSEP LAR Rev 0 Page E-24

Duke Energy Attachment E - Radioactive Release Transition NEI 04-02 Table E-1 Radioactive Release Compartment Review Compartment Fire Area/ Pre-Fire Plan Screened In? Evaluation Revisions Conclusion Fire Zone OMM-003 Section OTHER POTENTIAL RADIOLOGICAL RELEASE AREAS NOTE: The following areas and buildings are not included in the NFPA 805 Power Block as defined in Attachment I. These areas were reviewed as part of the analysis to ensure that all radiological areas with a potential for release were considered. These areas screened out of the analysis and were not added to the Power Block definition. Actions were taken to update Fire Pre-plans for guidance in these areas.

ISFSI, Incorporate into OMM-003, Fire Pre-Plans with standard statements concerning airborne contamination and water runoff.

ISFSI Service Building, Incorporate into OMM-003, Fire Pre-Plans with standard statements concerning airborne contamination and water runoff.

Building 230, Contaminated Storage Bldg., Incorporate into OMM-003, Fire Pre-Plans with standard statements concerning airborne contamination and water runoff.

Building 250, Outage Contaminated Storage Bldg., Incorporate into OMM-003, Fire Pre-Plans with standard statements concerning airborne contamination and water runoff.

Need to incorporate into the Pre-Fire Plan guidance for outside Yard areas where Radioactive Materials Areas (RMA's) and Sea-Land type cargo container storage may be present. (Implementation Item, Attachment S, Table S-3).

Page E-25 HBRSEPLARRevO HEIRSEP LAR Rev 0 Page E-25

Duke Energy Attachment F - Fire-induced MSOs Resolution F. Fire-Induced Multiple Spurious Operations Resolution 5 Pages Attached Page F-I HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page F-1

Duke Energy 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 HBRSEP MSOs of concern:

, Post-fire safe shutdown analysis

• HNP and CR3 MSO expert panel results

• Results of the previous HBRSEP MSO expert panel (RNP-0087, Rev. 0 -

September 2009)

• PWR Owners Group Generic MSO List (Table G-2 of NEI 00-01, Rev. 3)

• Internal events PRA insights The Robinson staff participants provided extensive Robinson experience and plant specific considerations. Following the initial expert panel of 2009, additional reviews were performed in 2012 of the updated PWROG generic MSO list associated with NEI 00-01, Revision 3.

The HBRSEP results are documented in:

" HBRSEP MSO Expert Panel Report (RNP-0087, Rev. 0), dated 9/9/2009

• HBRSEP MSO Expert Panel Report (RNP-0148, Rev. 0) dated 2/20/2013 Page F-2 HBRSEP LAR Rev 0 LAR Rev 0 Page F-2

Duke Energy Attachment F - Fire-induced MSOs Resolution Step 2 - Conduct an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. I 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.).

[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 2009 The initial expert panel was held on site at Robinson April 27 - May 1, 2009. The HBRSEP Expert Panel included individuals from both site and corporate engineering, plant Operations Department, corporate PRA, and industry consultants. The panel discussed fire-induced MSOs that could potentially impact fire safety. They also reviewed system flow diagrams to postulate and discuss new potential MSOs not previously considered. Documents used as guidance included:

• NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Rev. 1

• FAQ 07-0038 (Ref. ADAMS Accession Nos. ML072740262, ML073060506, ML082100034)

Training for the initial 2009 expert panel was conducted in the form of an introductory overview. Topics discussed included:

" Purpose and scope of the SSA

" PRA overview and results

• Overview training on the MSO issue

" Results of the Fire Testing (EPRI/NEI Testing), including:

o Likelihood of various spurious operation probabilities o Timing, including the likelihood that failures will occur close together in time, and issues affecting time to damage o Duration

" Overview of plant thermal hydraulic analyses that support safe shutdown.

HBRSEP LAR Rev 0 Page F-3

Due neq Attachment F - Fire-induced MSOs Resolution Duke Energy Attachment F Fire-Induced MSOs

- Resolution Additionally, the following key points were discussed:

" 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 would not be on individual fire area locations, but would focus on system/component interactions.

The Expert Panel started with a review of the scenarios that were developed from the following sources:

• Pressurized Water Reactor Owner's Group (PWROG) Generic MSO List (Working Draft Rev. E)

" Results of Harris and Crystal River Nuclear Plant MSO Expert Panels Additionally, the Expert Panel conducted "what if' discussions based on a review of system flow drawings to postulate/discuss new potential MSOs not previously considered.

The PWR Generic MSO List includes scenarios related to the following functions:

• Reactivity Control

• Reactor Coolant Makeup Control

" Reactor Coolant System Pressure Control

• Decay Heat Removal

• Support Systems

" Process Monitoring Using the PWR Generic MSO List as guidance, a step-by-step discussion was held, reviewing simplified flow diagrams, control diagrams, and system flow diagrams when necessary. The process also included postulating scenarios, discussing the potential consequences and likelihood of the scenario, discussing operator response, and recommending additional courses of action. Key considerations, in addition to potential consequences were:

• Whether the scenario of concern was currently modeled in the Robinson SSA,

• Whether procedures addressed the potential scenarios of concern, and

" Additional analyses or justification that may be necessary to document exclusion of a particular scenario Consensus was achieved in the expert panel process by discussing individual scenarios, reaching a conclusion, and asking for any dissenting opinions.

Page F-4 HBRSEP LAR Rev 0 LAR Rev 0 Page F-4

Duke Eneray Attachment F - Fire-induced MSOs Resolution Follow-on Expert Panel 2012 The follow-on expert panel meeting in May 2012 used the same information, but with an updated PWROG generic MSO list from NEI 00-01 Revision 3 and current lessons learned from the MSO process in FAQ 07-0038, rev. 3, and NFPA 805 pilot plants. As was done previously, training for the 2012 expert panel was conducted in the form of an introductory overview and included the topics discussed in 2009.

The follow-on meeting in 2012 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 (Working Draft Rev. E).

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 HBRSEP Results are documented in:

" NSCA- RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for Robinson Nuclear Plant

" Fire PRA- RNP-F/PSA-0066, Attachments 3 and 10 Page F-5 HBRSEPLARRevO HBRSEP LAR Rev 0 Page F-5

Duke Energy Attachment F - Fire-induced MSOs Resolution 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 (VFDR), 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.).

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 HBRSEP 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 HBRSEP Results are documented in:

" MSO Expert Panel Report (RNP-0087, Revision 0, HBRSEP MSO Expert Panel Report), September 2009

" MSO Expert Panel Report (RNP-0148, Revision 0, HBRSEP MSO Expert Panel Report), February 2013

" NSCA - RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H. B.

Robinson Nuclear Plant, Rev. 0

" Fire PRA - RNP-F/PSA-0066, rev. 5, Attachments 3, and 10

• HBRSEP Fire Safety Analyses, RNP-M/MECH-1 844 through 1876.

Page F-6 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page F-6

Duke Energy Attachment H - NEI 04-02 FACls 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 is revised to endorse a new revision of NEI 04-02 that incorporates the outstanding FAQs.

Page H-I HBRSEPLARRevO HBRSEP LAR Rev 0 Page H-1

Duke Energy 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 Utiliz ed 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 07-0038 Lessons learned on Multiple Spurious ML103090608 ML110140242 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 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page H-2

Duke Energy Attachment H - NEI 04-02 FAQs Summary Table Table H NEI 04-02 FAQs Utilized in LAR Submittal No. Rev. Title FAQ Ref. Closure Memo 08-0053 0 Kerite-FR Cable Failure Thresholds ML082660021 ML120060267 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 1 UFSAR Content ML121430035 ML121980557 12-0063 1 Fire Brigade Make-Up ML121670141 ML121980572 12-0064 1 Hot work/transient fire frequency: ML122550050 ML12346A488 influence factors 12-0067 1 Transformer Oil Collection Drain Basin ML13035A039 ML13037A425 Inspections

• Note: The FAQ submittal number was 08-0054 but the NRC closure memo for the FAQ was listed as 07-0054. FAQ 07-0054 was used to be consistent with the Closure Memo.

Page H-3 HBRSEP HBRSEP LAR Rev 00 LAR Rev Page H-3

Due neg Attachment I - Definition of Power Block Duke Energy Attachment I Definition of Power Block I. Definition of Power Block I Pages Attached Page I-I HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page 1-1

Duke Energy Attachment I- Definition of Power Block The structures in the Owner Controlled Area were evaluated to determine those structures that contain equipment that is required to meet the nuclear safety performance criteria and radioactive release performance criteria described in Section 1.5 of NFPA 805.

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 Fire Areas Notes Structures A1, A2, A3, A4, A5, A6, A7, A8, Includes Waste evaporator area Auxiliary Building A9, A10, All, A12, A13,C,A14, A15, A16, A17, A19, B, D,E, G4 Control Room A18 Includes Hagan Room Containment Building F G1 Includes the Dedicated Shutdown Transformer, Turbine Building Main and Auxiliary Start-Up Transformer Yard, Refueling Water Storage Tank, Primary Water Storage Tank, Condensate Storage Tank Diesel Fuel Oil Storage G2 Tank Intake Structure G3 G4 Includes the Cask Preparation Area, Hot Fuel Handling Building Machine Shop and Fuel Building, Radwaste Building, Purge inlet valve room, Diesel Fuel Oil Unloading/Transfer Area 115kV and 230kV G5 Switchyards Dedicated Shutdown G7 Diesel Generator Enclosure Residual Heat Removal H Pump Room YARD Includes Unit 1 and 2 Switchyard, Secondary Sampling Building, Nitrogen Storage and "C" Auxiliary Boiler, Condensate Polishing Building, YARD Deepwell Pump "A"Area South of Unit I Service Building, Deepwell Pump "C" area North side of O&M Building, and the Deepwell Pump "D" Enclosure, PAP East The Independent Spent Fuel Storage Installation Area was considered for Radioactive Release fire fighting activities. This area is not included in the NFPA 805 definition of power block or any analysis because it is licensed under 10 CFR Part 72.

Page 1-2 HBRSEP LAR LAR RevRev 00 Page 1-2

Duke Energy Attachment J - Fire Modeling V&V J. Fire Modeling V&V 21 Pages Attached Page J-1 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page J-1

Duke Energy Attachment J - Fire Modeling V&V INTRODUCTION This attachment documents the verification and validation (V&V) of the fire models as applied to the Robinson Fire PRA following the guidance documented in NUREG-1824, and NUREG-1934. These documents are relatively recent joint publications by the US NRC and the Electric Power Research Institute intended to provide guidance on how to conduct and document fire modeling studies, as well as develop the necessary V&V material for supporting these studies.

The analysis summarized in this appendix is based on the technical material documented in RNP-M/MECH-1884, Verification and Validation of Fire Models Supporting the Robinson Nuclear Plant (HBRSEP) Fire PRA. The summary covers all the fire models and the fire modeling applications within the Robinson Nuclear Plant Fire PRA as documented in the different calculations prepared for those purposes during the development of the Fire PRA. Each of the models used in the different calculations is identified and a V&V discussion is provided. The report also includes a summary table listing the fire models with the corresponding V&V results.

SCOPE The scope of this study includes the V&V of fire models based on the guidance available in NUREG-1824 and NUREG-1934 as applied in the HBRSEP Fire PRA. The following subsections list and describe the HBRSEP fire modeling calculations within the scope of the V&V study.

Zone Models (CFAST)

The computer model CFAST, is used in the main control room abandonment study documented in Report 0004-0042-412-002, Rev. 1, "Evaluation of Control Room Abandonment Times at the H. B. Robinson Nuclear Plant" and to document hot gas layer conditions in Fire Zone 20 in Report 0004-0042-000-001, Rev. 1, "Evaluation of the Development and Timing of Hot Gas Layer Conditions in HBRSEP Fire Zone 20",

respectively. The V&V for CFAST in the HBRSEP main control room abandonment study and in the Fire Zone 20 hot gas layer calculation is included in the respective report and is summarized in this document for completeness purposes.

Engineering Calculations (Hand Calculations)

The HBRSEP Fire PRA is characterized by a number of engineering (i.e., hand calculations) used throughout the analysis for various purposes. The following subsections provide a brief description of these calculations.

Hot Gas Layer Calculations The report RNP-M/MECH-1826, Rev. 1, "Hot Gas Layer Calculation", documents the approach for determining the damage time for cables immersed in a hot gas layer. The hand calculations used for this analysis are the MHQ room temperature correlation for rooms assuming an open door (NUREG-1805, Chapter 2.1) and the Beyler room temperature correlation for closed doors room (NUREG-1 805, Chapter 2.3). The document also includes an analysis for screening multi compartment combinations. In general, hot gas layer temperatures are calculated for selected fire zones. If the hot gas HBRSEP LAR Rev 0 Page J-2

Duke Energy Aftachment J - Fire Modeling V&V layer temperature is calculated to be lower than the damage thresholds for cables, the multi compartment scenario is screened.

Cable Tray Fire Propagation NED-M/MECH-1 009, Rev. 0, "Thermal Damage Time of Cables Above a Burning Ignition Source". This calculation describes the approach for determining the time to damage or ignition of the closest cable tray or conduit to an ignition source and subjected to fire plume conditions. The calculation produces a "look up" table for damage or ignition times that are used in the quantification process for calculating non suppression probabilities. The fire model within the scope of this validation and verification study is the Heskestad Plume Temperature Correlation documented in Chapter 9 of NUREG-1 805.

ZOI Calculations The ZOI calculations in the HBRSEP Fire PRA are based on hand calculations. These calculations are documented in the following reports:

NED-M/MECH-1008, "Fire Zone of Influence Calculation". The goal of this calculation is to calculate ZOI values for various fire sizes that are conservative, encompass a broad set of fuel packages, and integrate more effectively with the scoping fire modeling process. The fire models within the scope of this V&V study are the Heskestad Plume Temperature Correlation documented in Chapter 9 of NUREG-1 805 and the solid flame radiation model documented in Section 5.2 of NUREG-1805.

• NED-M/MECH-1 007, "Radiant Energy Target Damage Profile": The purpose of this document is to provide a refinement of the radiant energy ZOI calculation used for identification of transients from electrical cabinet fires. The fire models within the scope of this V&V and the solid flame radiation model documented in Section 5.2 of NUREG-1805.

" NED-M/MECH-1006, "Generic Fire Modeling Treatments": The generic treatments document offers a set of pre-defined ZOI calculations. A number of fire models are subjected to V&V. These models are listed in Table J-1.

• RNP-0206, "Analysis of Oil Fires for Compressors in the Lower Hallway (Fire Zone 7)". This report calculates damage that may occur in the vicinity of equipment (i.e., compressors) due to radiant affects and due to plume affects after an oil spill fire scenario. The report evaluates the fire size and determines whether resulting damage to cables and components takes place. The fire ZOI is determined using methods and tools provided by NUREG-1805, Fire Dynamics Tools.

Page J-3 HBRSEP LAR HBRSEP Rev 0 LAR Rev 0 Page J-3

Duke Energy Affachment J - Fire Modeling V&V REFERENCES This section lists the references utilized in this report to perform the fire model V&V.

References are classified as "Industry" and "plant specific".

Industry References

27. 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.

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

29. NUREG-1824, Volume 1, "V&V of Selected Fire Models for Nuclear Power Plant Applications Volume 1: Main Report," NUREG-1824 / EPRI 1011999, Salley, M. H, and Kassawara, R. P., NUREG-1824, Final Report, U.S.

Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C., May, 2007.

30. 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.

31. NUREG-1934, "Nuclear Power Plant Fire Modeling Application Guide,"

Salley, M. H. and Kassawara, R. P., NUREG-1934/EPRI-1019195, U.S.

Nuclear Regulatory Commission, Office of Nuclear Reactor Research, Washington, D. C., November, 2012.

32. 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.

33. NIST SP 1026, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) Technical Reference Guide," Jones, W. W., Peacock, R. D., Forney, G. P., and Reneke, P. A., National Institute of Standards and Technology, Gaithersburg, MD, April, 2009.

34. NIST SP 1041, "CFAST - Consolidated Model of Fire Growth and Smoke Transport (Version 6) User's Guide," Peacock, R. D., Jones, W. W., Reneke, P. A., and Forney, G. P., National Institute of Standards and Technology, Gaithersburg, MD, December, 2008.

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

Peacock, R. D., McGrattan, K., Klein, B., Jones, W. W., and Reneke, P. A.,

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

HBRSEP LAR Rev 0 Page J-4

Duke Energy Aftachment J - Fire Modeling V&V

36. NRL/MR/6180-04-8746, "Verification and Validation Final Report for Fire and Smoke Spread Modeling and Simulation Support of T-AKE Test and Evaluation," Tatem, P.A., Budnick, E.K., Hunt, S.P., Trelles, J., Scheffey, J.L.,

White, D.A., Bailey, J., Hoover, J., and Williams, F.W., Naval Research Laboratory, Washington, DC, 2004.

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

38. 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.

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

40. Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current," Report Number 34, Building Research Institute, Tokyo, Japan, 1960.

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

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

43. SFPE Handbook of Fire Protection Engineering, Section 3-1, "Heat Release Rates," Babrauskas, V., The SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.

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

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

46. 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-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.

Page J-5 HBRSEP HBRSEP LAR Rev 0 LAR Rev 0 Page J-5

Duke Energy Aftachment J - Fire Modelinq V&V Plant Specific References

60. RNP-M/MECH-1 826, "Hot Gas Layer Calculation".

61. NED-M/MECH-1008, "Fire Zone of Influence Calculation".

62. NED-M/MECH-1009, "Thermal Damage Time of Cables Above a Burning Ignition Source".

63. NED-M/MECH-1006, "Generic Fire Modeling Treatments".

64. NED-M/MECH-1007, "Radiant Energy Target Damage Profile".

65. Report Number 0004-0042-412-002, "Evaluation of Main Control Room Abandonment Times at the H.B. Robinson Nuclear Plant".

66. Report Number 0004-0042-000-001, "Evaluation of the Development and Timing of Hot Gas Layer Conditions in RNP Fire Zone 20".

67. RNP-0206, Analysis of Oil Fires for Compressors in the Lower Hallway (Fire Zone 7).

VERIFICATION AND VALIDATION This section includes Table J-1 and Table J-2, which present a summary of the fire models with the corresponding V&V results. Specifically, Table J-1 summarizes the verification and validation results for the different fire modeling calculations listed earlier under the scope section. Table J-2 is specifically devoted to discussing the validation for the fire models used in the generic fire modeling treatment document. The technical material supporting the summary presented in these tables is documented in RNP-M/MECH-1884, Verification and Validation of Fire Models Supporting the Robinson Nuclear Plant (HBRSEP) Fire PRA.

Page J-6 HBRSEP LAR Rev 0 LAR Rev 0 Page J-6

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation i I The ZOI calculations for the Fire Froude number present a few "out of range" results. All of the "out of range" cases are due to calculations exceeding the upper limit of the range, suggesting high intensity fires for the selected fire diameter. One reason for exceeding the upper limit is the use of the 98th percentile heat release rates for the corresponding fire diameters. Based on the guidance in Chapter 8 of NUREG, 98th percentile heat release rate values are used for screening and can be The verification of the models Heskestad considered on the high end of the Use of U.S. Nuclear used in support of calculation Plume values assigned to ignition sources. In Regulatory NED-M/MECH-1008, is Temperature addition, setting the Froude number Commission (NRC) provided in NUREG-1805, Correlation calculation to the upper range limit of Fire Dynamics Tools which contains pre-documented in 2.4 for the 98th percentile heat release (FDTs) [NUREG- programmed Microsoft Excel NED-M/MECH-1008, Chapter 9 of rate values would result in a larger 5.1 1805.0] to determine Spreadsheets. The Fire ZOI Calculations NUREG-1805. diameter. With a larger diameter, the the ZOI of a fire spreadsheets from NUREG Solid flame flame height calculation would result in scenario in support of 1805 are used directly in radiation model shorter flame lengths, and plume scenario development NED-M/MECH-1008, documented in temperature calculations would suggest for the HBRSEP Fire (Attachment 1) and therefore Section 5.2 of lower temperatures. The "out of range" PRA additional verification is not NUREG-1805. results are based on conservative ZOI needed.

calculations for the Fire PRA.

Parameters are "in range" for the fire plume application.

Parameters are "out of range" for the use of the solid flame radiation model.

The reason for number of ZOI results are "out of range" is because the ZOI distances are close to the flames and the experiments selected for validation purposes measured radiation at longer distances from the flames. This is a limitation on the available data for validation and not necessarily a HBRSEP LAR Rev 0 Page J-7

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation limitation on the use of the solid flame radiation model for calculating horizontal components of the ZOI for Fire PRA applications. To account for this limitation, it is noted that validation results from Figure 6-8 in Volume 3 of NUREG-1824 suggest significant heat flux over predictions over the intensity levels used for ZOI calculations (i.e.,

between 6 and 11 kW/m 2) that would result in longer, and therefore conservative, horizontal distances.

The Fire Froude Number for all the cases in the vertical ZOI calculations are within the validation range, indicating the heat release rates relative to the fire diameters for the scenarios described by calculation RNP-0206, are Use of U.S. Nuclear The verification of the models within the scope of NUREG-1824. With Regulatory Heskestad used in support of calculation regards to the flame length ratio Commission (NRC) Plume NED-M/MECH-1008, is dimensionless parameter, the majority Fire Dynamics Tools Temperature provided in NUREG-1805, of the unconfined cases are above the RNP-0206, Analysis (FDTs)

[NUREG-1805.0] to Correlation documented in which contains Microsoft pre-programmed heightrange, valid an the exceeds indication thatofthe distance theflame of Oil Fires for determine a fire ZOI for Chapter 9 of Excel Spreadsheets. The target above the fire for both 5.2 tre bv h iefrbt Lower Hallwayin the Compressors two pieces of NUREG-1805. spreadsheets from compressors. These are conservative Lower Hallway equipment, the Station Solid flame NUREG-1805 are used calculations for the Fire PRA given that (Fire Zone 7) Air Compressor Motor, radiation model directly in the thermoplastic cables would be in and the Instrument Air documented in NED-M/MECH-1008, direct contact with the flame and could Compressor "B" for Section 5.2 of (Attachment 1) and therefore ignite. That is if even with the "out of unconfined and NUREG-1805. additional verification is not range results" the cables are set to fail confined fire scenarios. needed, in the fire PRA quantification The confined fire scenario calculation for both compressors results in Fire Froude number and a flame length ratio within the validation ranges included in NUREG-1824.

The Fire Froude numbers calculated for HIBRSEP LAR Rev 0 Page J-8

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation both the unconfined and confined cases for the horizontal ZOI are within the validation range reported by NUREG-1824.

Most of the radial distance ratio parameters for unconfined fire scenarios are below the valid range for this parameter, indicating that the ZOI for these fire scenarios is characterized by distances close to the flames. These calculations were conducted with the solid flame radiation model described in Chapter 5.2 of NUREG-1805. A review of Figure 6-8 in Volume 3 of NUREG-1824 suggests that the majority of the validation results (with a few exceptions for Cable G in radiation ranges larger than the 6 and 11 kW/m for ZOI calculations) over predict flame radiation, which result in longer horizontal distances for the ZOI. This is a limitation on the available data for validation and not necessarily a limitation on the use of the solid flame radiation model for calculating horizontal components of the ZOI for Fire PRA applications.

The confined fire scenario calculation for both compressors for fire scenarios 2 and 3 results in a Fire Froude number and a radial distance ratio parameter within the validation ranges included in the NUREG-1824 NED-M/MECH-1009, Calculation NED- Heskestad The Calculation Several Fire Froude Numbers fall "out M/MECH-1009, Plume NED-M/MECH-1009, was of range" either under or over the valid Thermal Damage 5.3 determines the time at Temperature developed under a QA range. All of the "out of range" cases Time of Cables Above ofaburnn which damage occurs Correlation program. During the design exceeding the upper limit of the range, to cables suspended documented in verification review, an suggest high intensity fires for the HBIRSEP LAR Rev 0 Page J-9

Duke Energy Attachment J - Fire Modelina V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation Ignition Source over a burning Chapter 9 of independent check of the selected fire diameter. One reason for electrical cabinet. NUREG-1805 quantitative results was exceeding the upper limit is the use of performed, therefore, results the 9 8th percentile heat release rates for were found to be consistent the corresponding fire diameters.

and the calculations included Based on the guidance in Chapter 8 of in the report have been NUREG, 98h percentile heat release verified. rate values are used for screening and can be considered on the high end of the values assigned to ignition sources.

In addition, setting the Froude number calculation to the upper range limit of 2.4 for the 9 8 th percentile heat release rate values would result in a larger diameter. With a larger diameter, the flame height calculation would result in shorter flame lengths, and plume temperature calculations would suggest lower temperatures. The "out of range" results are based on conservative ZOI calculations for the Fire PRA.

Cases where the Fire Froude Number results below the validation range indicate low intensity fires where the fire HRR is low compared to the pool fire area. This occurs for the lowest three HRR cases included in the calculation (i.e., 69, 143 and 211 kW) at different fire diameters. These are cases where the thermal plume that is expected from the ignition source fire could be wider than the range evaluated in NUREG-1824. A wider thermal plume will have a greater entrainment rate than one associated with a similar heat release rate fire that has a smaller diameter. This means that the conditions relative to a source fire that falls within the validation range will be less severe in terms of temperature.

Page J-10 HBRSEP LAR Rev 00 Page J-10

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation These are cases where damage to the target occurs most likely due to direct flame impingement before damage time to the target occurs.

Several flame length ratio calculations resulted "out of range". All of these results are above the high end of the validation range, meaning that the length of the flame was always greater than the height of the target above the fire source causing direct flame impingement on the target. For these cases, given the diameter and HRR of the fire, direct flame impingement occurred to the target cable. Thus, a larger fire diameter would result in flame length ratios within the validation range but lower Fire Froude Numbers. Thus, the values of flame length ratio that are not within the validation range are based on conservative calculations for the Fire PRA.

To ensure the equations were The comparison of dimensionless Calculation NED- coded correctly in the parameters with the validation range M/MECH-1007, utilizes spreadsheets used in the suggest a number of "out of range" the U.S. Nuclear calculation, the spreadsheet results, which are expected for both the results were checked against Fire Froude Number and the radial Regulatory Solid flame the results of the distance ratio dimensionless parameter.

NED-M/MECH-1007, Commission (NRC) radiation model NUREG-1805 FDTs Solid For the radial distance ratio Radiant Energy 5.4 Fire Dynamics Tools documented in Flame Model 2 spreadsheet dimensionless parameter, all the Target Damage (FDTs) [NUREG- Section 5.2 of for identical inputs. Both calculations that are "out of range" are Profile 1805.0] to determine a NUREG-1805. spreadsheet models were on the low side of the range. This radiative electrical ZOI fromfires foundst pode produce tese happens because cabinet eletrialcabne fiesresults found to(NED-M/MECH-1007),

the same tefae the flames and h the n the target is close to xeiet experiments to qualified and reeuts spreaheets selected for validation purposes unqualified* cables.

able.

unualiied .*therefore used in thethecalculations spreadsheetsare measured radiation at longer distances considered verified, of the target from the flames. This is a limitation on the available data for HBRSEP LAR Rev 0 Page J-11I

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation validation and not necessarily a limitation on the use of the solid flame radiation model for calculating horizontal components of the ZOI for Fire PRA applications. To account for this limitation, it is noted that validation results, from Figure 6-8 in Volume 3 of NUREG-1 824, suggest significant heat flux over predictions over the intensity levels used for ZOI calculations (i.e.,

between 6 and 11 kW/m2) that would result in longer, and therefore conservative, horizontal distances.

Calculation The Beyler room temperature RNP-M/MECH-1826, correlation was developed using data determines the fire heat The hand with a maximum temperature rise of release rate necessary calculations 150°C. Extrapolation of this correlation to generate a damaging used for this to higher temperatures (330*C) is hot gas layer within a analysis are the justified by using the Beyler correlation hot colayertethn o M s aro tonly when it is the most conservative compartment or MHQ room The fire modeling documented result (i.e., lower estimate of HRR for multicompartment for a temperature in this calculation is a room-wide damage to cables),

givenformore,tsreioms f Microsoft Excel Spreadsheet compared to the MQH correlation,

• Furthermore, this rooms RNP-M/MECH-1826, calculation describes assuming an supplemented with VBA which is validated at higher HoPGs aer 5.5 the process for open door Macros. The spreadsheet is a temperatures Hot Gas Layer 5.5 crediting the "heat (NUREG-1805, custom built fire modeling tool The results show that the majority of the Calculation soak" time. The "heat Chapter 2.1) that uses the same closed- compartment ratio parameters are soak" time refers to the and the Beyler form room temperature within the valid range, suggesting that lag time between the room correlations (Sections 5.1 and the room size of these fire scenarios temperature temperature 5.3 of NUREG-1805) that are was included in the V&V study surrounding the cable correlation for provided in NUREG-1824 described in NUREG-1824. Those targets and the closed doors compartment aspect ratios that fall temperatures inside the room outside the application range do so on cable targets (NUREG-1805, both ends of the range. This can be generating the Chapter 2.3). explained by the limited experiments electrical damage selected for the validation study. As and/or ignition. indicated in NUREG-.1934, the selected experiments are representative of HBRSEP LAR Rev 0 Page J-1 2

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models In Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation various types of spaces in commercial NPPs, but do not encompass all possible geometries or applications.

This is a limitation on the available data for validation and not necessarily a limitation on the use of the model for calculating HGL scenarios applicable to the Fire PRA. To address this limitation, it is noted that both the MQH and Beyler room temperature models are reported to overpredict room temperatures for most configurations in Table 3-1 of NUREG-1824, Volume 1 (which lists a yellow-plus) and Table 4-1 in NUREG-1934 (which suggests an average bias of 1.44). This over prediction throughout the evaluated scenarios suggest that the configurations that are outside the validation range in this application will also result in temperature over predictions.

The "Generic Fire Modeling Treatments," The calculation development Revision 0 document is and review process in place at used to establish ZOI the time the "Generic Fire NED-M/MECH-1006, for specific classes of Listed in Table Modeling Treatments" Generic Fire 5.6 ignition sources and 2 later in this document was prepared Listed in Table 2 later in this section.

Modeling Treatments primarily serves as a section included contributions from a screening calculation in calculation preparer, a the Fire PRA under calculation reviewer, and a NUREG/CR-6850 calculation approver.

Sections 8 and 11.

Report No. Calculation of main Attachment 4 of Report A full validation study for the analysis is 0004-0042-412-002, control room CFAST, 0004-0042-412-002, includes described in Section A4.5.1 of Report Evaluation of Control 5.7 abandonment times. Version 6.1.1 a software description and 0004-0042-412-002. The Room Abandonment The abandonment benchmark V&V. The non-dimensional parameters that affect Times at the H. B. times are then used as attachment provides a the model results as documented in HBRSEP LAR Rev 0 Page J-1 3

Duke Energy Attachment J - Fire Modelina V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation Robinson Nuclear inputs to the risk description of the verification NUREG-1824, Volumes 1 and 5 and Plant quantification of main documentation for CFAST. NUREG-1 934, include the model control room fire The primary documents geometry, the equivalence ratio, the fire scenarios. The report applicable to this effort are Froude Number, and the flame length provides operator NIST SP 1086 (Ref. 9), ratio.

abandonment times in NUREG-1824, Volume 1 (Ref. All non-dimensional geometry the HBRSEP MCR due 3), and NUREG-1824, Volume parameters fall within the to visibility reduction 5 (Ref. 20). A benchmark NUREG-1824, Vol. 1 validation range of and/or temperature installation and verification 0.6-5.7.

increase for fire procedure is provided by NIST scenarios in the (Ref. 8) to ensure correct Table A4-5 of Report 0004-0042-412-HBRSEP MCR. Fire installation and proper function 002, shows the approximate Fire sizes are postulated of the CFAST model Froude Number for NUREG/CR-6850, using the discretized components. This procedure Appendix E Case 8 (Transient Fires).

distributions for specific was performed as part of the The table indicates that the Fire Froude types of electronic verification process. Number falls below the NUREG-1824 equipment fires and validation range of 0.4 - 2.4 in nearly transient combustible all cases, which means that the thermal fires as described in plume that is expected from the ignition NUREG/CR-6850. source fire could be wider than the range evaluated in NUREG-1824. A wider thermal plume will have a greater entrainment rate than one associated with a similar heat release rate fire that has a smaller diameter. This means that the conditions relative to a source fire that falls within the validation range will be less severe both in terms of the concentration of combustion products and the temperature. In the case of the Main Control Room the results are conservative when applied to low Fire Froude Number fire scenarios.

Regarding the equivalence ratio calculation, Report 0004-0042-412-002, demonstrates that all cases fall within the NUREG-1 824 validation range (i.e.,

0.04-0.6). The global equivalence ratio for normal air supply to the MCR is HBRSEP LAR Rev 0 Page J-14

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models In Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation assessed using the ratio of the maximum supported fire size to the fire size postulated. The report indicates that the fresh air supply and the initial oxygen concentration within the HBRSEP MCR is capable of supporting a fire on the order of 1,990 kW (1890 Btu/s) for twenty-five minutes when the HVAC system is providing outside air and 1,440 kW (1,360 kW) when the HVAC system is not providing outside air. At an equivalence ratio of 0.6, the maximum fire size would be about 1,194 kW (1,132 Btu/s). This is larger than the transient, single bundle electrical panel, and multiple bundle electrical panel fire scenarios.

However, the workstation and the propagating MCB panel fire scenarios (Bin 7 and above) have fire sizes that are greater than 1,194 kW (1,132 Btu/s) at least for a portion of the scenario.

When the time at which the abandonment is predicted is factored into the equivalence ratio calculation, it is shown that all cases have equivalence ratios less than 0.6 up to the time that abandonment is predicted.

Analysis of the hot gas Attachment B of Report Tables A2-7 through A2-10 of Report Report No. layer temperature and 0004-0042-000-001, describes 0004-0042-000-001, summarize the 0004-0042-000-001, soot concentration the Verification for the CFAST non-dimensional parameters for a Evaluation of the conditions in the model Version 6.1.1.54. The transient fuel package fire and the Development and 5.8 Robinson Nuclear Plant CFAST Version attachment provides a electrical panel fires as located in the Timing of the Hot (HBRSEP) fire 6.1.1.54 description of the verification El/E2 Switchgear Room and the Gas Layer Conditions compartments for documentation for CFAST. Safeguards Room for the initial ignition in HBRSEP Fire transient ignition The primary documents source. The tables indicate that the Zone 20 sources and electronic applicable to this effort are transient fire scenario parameters and panel ignition sources . NIST SP 1086 (Ref. 9), most of the panel fire scenario HEIRSEP LAR Rev 0 Page J-156

Duke Energy Aftachment J - Fire Modelina V&V Table J-1: Summary of V & V Results for Fire Models in Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation that involve secondary NUREG-1824, Volume 1 (Ref. parameters fall within the combustibles (cable 3), and NUREG-1824, Volume NUREG-1824, Volume 1 parameter trays). Fire scenarios 5 (Ref. 20). space range. Several of the panel fires are evaluated in the have fire Froude Numbers that are E1/E2 Switchgear somewhat below or above the Room and the NUREG-1824, Volume 1 (Ref. 3) range.

Safeguards Room, in In addition, the flame height to the Reactor Auxiliary enclosure ratio is greater than the Building (RAB). NUREG, Vol. 1 (Ref. 3) range for the smallest diameter panel (MCCs) in the Safeguards Room.

The large flame length predicted for the Safeguards Room panel fire scenario is not expected to adversely affect the calculation results. In the case of the Halon actuation, the fire size at the time the smoke detectors actuate is much smaller than the peak fire size upon which the values in Table A2-8 are derived. In this case, the flame length will be shorter than the ceiling and the application will be within the NUREG-1824, Volume 1 (Ref. 3) range.

Although this line of reasoning does not apply to the hot gas layer temperature calculation, it may be inferred that the overall treatment of the electronic panel fire, especially in the MCCs, is not representative of the way in which they will actually behave if ignited. The MCCs are relatively well sealed and external combustion may occur at gaps or seams in the MCC enclosure. The CFAST model conservatively removes the metal enclosure and places the fire 0.3 m (1 ft) below the panel top.

The actual flame length will be shorter and within the NUREG-1824, Vol. 1 Page J-16 HBRSEP LAR RevRev 0 0 Page J-1 6

Duke Energy Attachment J - Fire Modeling V&V Table J-1: Summary of V & V Results for Fire Models In Specific HBRSEP Fire PRA Applications Report Calculation Section Application Fire Models Verification Validation (Ref. 3) test basis.

The fire Froude Numbers for several panels are shown in Tables A2-7 and A2-8 of Report 0004-0042-000-001 to be either lower or higher than the range evaluated in NUREG-1824, Volume 1 (Ref. 3). A similar argument that was developed for the flame length applies to the fire Froude Number. The characteristic length is somewhat arbitrarily established using the panel plan dimensions. The actual characteristic length is difficult to assess since the panel is not a simple two-dimensional fuel package. Burning will occur at the vents, if present, and gaps and seams, all of which are smaller than the characteristic dimension. From a macroscopic perspective, the 464 kW (440 Btu/s) electronic panel fire is a common plant ignition source and is not significantly different from the types of source fires considered in the NUREG-1 824, Volume 1 (Ref. 3) test series.

Page J-17 HBRSEP LAR Rev 00 LAR Rev Page J-17

Duke Energy Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations.

Location in Reference Generic Fire in "Generic Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Mdeaent" Verification* (Ref. 11)

(Ref. 11) Treatments" Document*

Flame Height Page 18 Heskestad Provides a limit Directly (Ref. 19);

Heskestad

)5 (Ref. 20) on the use of the ZO

-g

< loglo[\K cpToo

(,),/

. Q]

5-1 55 NUREG-1824, Volume 3 (Ref.

23) < 3000 10 4rAH Indirectly_

In practice, wood and hydrocarbon NUREG-1824, fuels, momentum or buoyancy Volume 5 (Ref.

dominated, with diameters between 42) 0.05 - 10 m (0.16 - 33 ft). (Correlation used in CFAST)

Point Source Page 19 Modak (Ref. Lateral extent of Isotropic flame radiation. Compared NUREG-1824, Predicted heat flux at Model 45) ZOI - with data for 0.37 m (1.2 ft) diameter Volume 3 (Ref. target is less than 5 comparison to PMMA pool fire and a target located 23); kW/m2 (0.4 4 Btu/s-ft2 )

other methods at a R ratio of 10. SFPE (Ref. 24) per SFPE.

R Method of Page 19 Shokri et al. Lateral extent of Pool aspect ratio less than 2.5. SFPE (Ref. 24) Ground based vertical Shokri and (Ref. 46) ZOI - Hydrocarbon fuel in pools with a NUREG-1824, target.

Beyler comparison to diameter between 1 - 30 m (3.3 - Volume 3 (Ref.

other methods 98 ft). 23)

Vertical target, ground level.

Method of Page 20 Mudan (Ref. Lateral extent of Round pools; SFPE (Ref. 24) Total energy emitted by Mudan (and 47) 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-18 HBRSEP LAR Rev 00 Page J-1 8

Duke Energy Attachment J - Fire Modelinq V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations.

Location in Reference in "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Treling Verification* (Ref. 11)

(Ref. 11) Document*

Method of Page 20 Shokri et al. Lateral extent of Round pools; SFPE (Ref. 24) Predicted heat flux at Shokri and (Ref. 46) ZOI Hydrocarbon fuel in pools with a target is greater than 5 Beyler diameter between 1 - 50 m (3.3 - NUREG-1824, kWIm 2 (0.44 Btu/s-ft2 )

164 ft). Volume 3 per SFPE (Ref. 24).

(Ref. 23) Shown to produce most conservative heat flux over range of scenarios considered among all methods considered.

Plume heat Page 22 Wakamatsu Vertical extent of Fires with an aspect ratio of about 1 Wakamatsu et Area source fires with fluxes et al. (Ref. ZOI and having a plan area less than 1 al. (Ref. 48) aspect ratio - 1. Used

48) m, (0.09 ft2). (larger fires) with plume centerline SFPE Handbook temperature correlation; of Fire most severe of the two is Protection used as basis for the ZOI Engineering, dimension. This is not a Section 2-14 constraint in the fire (Ref. 49) model analysis for the cases evaluated.

Plume Page 23 Yokoi (Ref. Vertical extent of Alcohol lamp assumed to effectively NUREG-1824, Area source fires with centerline 21); ZOI be a fire with a diameter -0.1 m Volume 3 aspect ratio - 1. Used temperature Beyler (Ref. (0.33 ft). (Ref. 23); with plume flux

50) SFPE Handbook correlation; most severe of Fire of the two is used as Protection basis for the ZOI Engineering, dimension.

Section 2-1 (Ref. 51)

Page J-19 HBRSEPLARRevO HBRSEP LAR Rev 0 Page J-19

Duke Energy Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations.

Location in Reference Generic Fire in "Generic Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Treaent" Verification* (Ref. 11)

(Ref. 11) Treatments" Document*

Hydrocarbon Page 51 SFPE Determine heat Hydrocarbon spill fires on concrete None. Based on None. Transition from spill fire size Handbook of release rate for surfaces ranging from -1 to -10 m limited number unconfined spill fire to Fire unconfined (3.3 - 33 ft) in diameter, of observations, deep pool burning Protection hydrocarbon assumed to be abrupt.

Engineering, spill fires.

Section 2-15 (Ref. 52)

Flame Page 100 SFPE Determine the Corner fires ranging from -10 to None. Based on None. Offset is assumed extension Handbook of fire offset for -1,000 kW (9.5 - 948 Btu/s). Fires limited number equal to the depth of the Fire open panel fires. included gas burners and of observations, ceiling jet from the Protection hydrocarbon pans. experiments.

Engineering, Section 2-14 (Ref. 53)

Line source Page 101 Delichatsios Determine the Theoretical development. SFPE Handbook None. Transition to area flame height (Ref. 54) vertical extent of of Fire source assumed for the ZOI Protection aspect plan ratios less Engineering, than four. Maximum of Section 2-14 area and line source (Ref. 49) predictions used in this region.

Corner flame Page 108 SFPE Determine the Corner fires ranging from -10 to None. None.

height Handbook of vertical extent of -1,000 kW (9.5 - 948 Btu/s). Fires Correlation form Fire the ZOI included gas burners and is consistent with Protection hydrocarbon pans. other methods; Engineering, comparison to Section 2-14 dataset from (Ref. 53) SFPE Handbook, Section 2-14 (Ref. 53) provides basis.

HBRSEP LAR Rev 0 Page J-20

Duke Energy Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations.

Location in inReference "Generic Generic Fire Fire Subsequent Limits in Generic Fire Correlation Modeling Modeling Application Original Correlation Range Validation and Modeling Treatments Treatments Treatments" Verification* (Ref. 11)

(Ref. 11) Document*

Air mass flow Page 140 Kawagoe Compare Small scale, % scale, and full scale Drysdale None. SFPE (Ref. 57) through (Ref. 55) mechanical single rooms with concrete and steel (Ref. 56); spaces with a wide range opening ventilation and boundaries. Vent sizes and thus SFPE (Ref. 57) of opening factors.

natural opening factor varied. Wood crib ventilation fuels.

Line fire flame Page 210 Yuan et al. Provides a limit None. None.

height (Ref. 22) on the use of the Z Correlation form ZOI (ZOI); 0.002 < T < 0.6 is consistent with Extent of ZOI for other methods; cable tray fires. In practice, from the base to several comparison to times the flame height based on dataset from 0.015 - 0.05 m (0.05 - 0.16 ft) wide Yuan et al.

gas burners. (Ref. 22) provides basis.

Cable heat Page 210 NBSIR 85- Provides Cables with heat release rates per None. Correlation predicts a release rate 3196 (Ref. assurance that unit area ranging from about 100 - lower heat release rate per unit area 58) the method used 1,000 kW/m 2 (8.8 - 88 Btu/s-ft 2). than assumed in the is bounding Treatments and is based on test data.

Line fire plume Page 212 Yuan et al. Provides a limit None. None.

centerline (Ref. 22) on the use of the Z Correlation form temperature ZOI (ZOI); 0.002 < T < 0.6 is consistent with Extent of ZOI for other methods; cable tray fires. In practice, from the base to several comparison to times the flame height based on dataset from 0.015 - 0.05 m (0.05 - 0.16 ft) wide Yuan et al.

gas burners. (Ref. 22) provides basis.

Page J-21 HBRSEP HBRSEP LAR Rev 0 LAR Rev 0 Page J-21

Duke Energy Attachment J - Fire Modeling V&V Table J-2: V & V Basis for Fire Models I Model Correlations Used: Generic Fire Modeling Treatments Correlations.

Location in Reference Generic Fire in "Generic Subsequent Limits in Generic Fire Correlation Modeling Fire Application Original Correlation Range Validation and Modeling Treatments Treatments Modeling Verification* (Ref. 11)

(Ref. 11) Treatments" Document*

Ventilation Page 283 Babrauskas Assessing the Ventilation factors between 0.06 - SFPE (Ref. 57) None. Provides depth in limited fire size (Ref. 59) significance of 7.51. the analysis of the vent position on Fire sizes between 11 - 2,800 kW selected vent positions.

the hot gas layer (10 - 2,654 Btu/s) The global equivalence temperature Wood, plastic, and natural gas fuels. ratio provides an alternate measure of the applicability of the analysis and for reported output is within the validation range of CFAST.

HBRSEP LAR Rev 0 Page J-22

Duke Energy Attachment K - Existing Licensing Action Transition K. Existing Licensing Action Transition 15 Pages Attached Page K-I HBRSEP LAR HBRSEP Rev 00 LAR Rev Page K-1

Duke Energy Attachment K - Existing Licensing Action Transition Licensing Actions Licensing Area C Exemption to Section III.G.2 of Appendix R to 10 CFR Part 50 (10-17-90)

Action Licensing The exemption from the requirements of Section III.G.2 of Appendix R of 10CFR50 allows additional Basis intervening cable combustibles to be installed in the CCW Pump Room. The previous configuration of the cable trays in the CCW Pump Room was approved in an exemption granted 10/25/1984.

The licensee has provided information on the level of fire protection already provided. These provisions include:

1. An early warning redundant cross-zoned fire detection system is provided;

2. A partial fire suppression sprinkler system is installed above all cable trays (including the presently proposed cable trays) and pumps;

3. No other combustible materials are in the CCW pump room except for one quart of lubricating oil with a high flashpoint (350'F) for each of the three pumps;

4. A portable fire extinguisher is in the CCW pump room, and additional fire extinguishers and manual fire hose stations are in adjacent areas;

5. Power cables to the redundant CCW pumps are installed in conduits with a 1-hour fire barrier wrap. All other cables are coated with flame retardant material or are IEEE-383, Vertical Flame Test, qualified. All cable trays (including the proposed cable trays) are qualified to IEEE 383.

Based on our review of the above information, the staff concludes that the licensee's existing fire protection configuration provides an equivalent level of safety to that achieved by compliance with Appendix R to 10 CFR Part 50.

Therefore, the Commission has determined that the an exemption from Section IIl.G.2.b of Appendix R to 10 CFR Part 50 in the CCW pump room should be granted.

Licensing 10/17/90 Transitioned El Date Reference Document Doc Detail NLS-90-148, CCW Pump Room Exemption, 7/30/90 NLS-90-166, Additional Information Concerning Component Cooling Water (CCW) Pump Room Exemption Request, 8/16/90 NLU-84-687, H. B. Robinson Steam Electric Plant Unit 2 (HBR-2) Fire Protection Appendix R to 10 CFR PART 50, Items III.G.2., 10/25/84 NRC-90-622, Exemption from Requirements of Section III.G.2 of Appendix R of 10CFR Part 50 - H. B. Robinson Steam Electric Plant No. 2, 10/17/90 HBRSEP LAR Rev 0 Page K-2

Duke Energy Attachment K - Existing Licensing Action Transition Licensing Area C Exemption to Section III.G.2 of Appendix R to 10 CFR Part 50 (10-25-84)

Action Licensing The Commission issued an Exemption to certain requirements of Appendix R to 10 CFR Part 50 in Basis response to CP&L letter dated March 16, 1982, as supplemented by letters dated April 27, 1982 and April 25, 1984. The exemption pertains to fire protection for the component cooling water pump room.

Specifically, HBRSEP is exempt from Section III.G.2 of Appendix R to 10 CFR Part 50 for the component cooling water pump room to the extent that it requires complete area-wide automatic fire suppression.

The exemption is conditioned on implementing fire protection improvements in this area as proposed.

These improvements must be completed in accordance with the provisions of 10 CFR 50.48(c). The basis for the Exemption is:

The CCW pump room is located in the Auxiliary Building on elevation 24600". The area is separated from other plant areas by 3-hour fire rated barriers. Fire protection features in the area consist of early warning smoke detectors, manual hose stations and portable hose stations.

The component cooling water pumps are located on a north-south orientation. Of the end pumps, pump A is redundant to pump C. Pumps A and C are 24 ft from centerline to centerline with approximately 20 ft (19' 11") between the closest points of each pump. Only one of the three (3) pumps and one (1) component cooling heat exchanger are required for safe shutdown.

By letter dated April 25, 1984, the licensee proposed to provide a one-hour barrier around the power supply cables to both pumps A and C, to enhance the protection of the component cooling water pump room and preclude ignition of the power cables to both pumps A and C. In addition, a partial fire suppression system will be installed in the area of the component cooling water pumps and the aisle in front of the pumps.

The partial automatic sprinkler system will be installed to protect the CCW pumps and associated cabling.

The combustible loading in the remaining areas of the CCW pump room is low and fire hazards do not exit.

If a fire occurred in the non-sprinklered section, the early warning smoke detectors would alert the fire brigade. The partial sprinkler system would prevent a fire from damaging redundant CCW components until the fire brigade could manually extinguish the fire.

It is the staffs opinion that the configuration of the area, combined with the response of the fire brigade, would prevent a fire from growing to a size which would overwhelm the partial sprinkler system. The addition of more sprinkler coverage would not significantly enhance safety. The existing protection with the proposed modifications will provide reasonable assurance that one safe shutdown division will be free of fire damage and will achieve an acceptable level of fire protection equivalent to that provided by Section III.G.2.

Therefore, the licensee's request for exemption for the CCW pump room should be granted.

Licensing 10/25/84 Transitioned El Date Reference Document Doc Detail LAP-83-210, Additional Information Concerning Pending Exemption Requests -

Appendix R, 617/83 MAR1682, Conformance to the Requirements of 10CFR50, Appendix R, Section III.G and Response to Generic Letter 81-12, 3/16/82 NLS-84-171, CCW Pump Room Exemption Request, 4/25/84 HBRSEP LAR Rev 0 Page K-3

Duke Energy Attachment K - Existing Licensing Action Transition NLU-84-687, H. B. Robinson Steam Electric Plant Unit 2 (HBR-2) Fire Protection Appendix R to 10 CFR PART 50, Items III.G.2., 10/25/84 RCIA-2, Appendix R Exemption Request, 4/27/82 Licensing Actions Licensing Area F Exemption from the Requirements of Section III.G.2.f of Appendix R to 10 CFR 50 Action Licensing Note: The requirement from NFPA 805 for a radiant energy shield is for a 1/2 hour fire barrier. The NRC Basis documented that the cable is rated for 93 minute fire exposure.

By letter dated August 17, 1984, the licensee requested an exemption from Section III.G.2.f to the extent that it would require the Rockbestos cables inside containment be kept "free of fire damage." This submittal also provided technical information previously requested by the NRC staff.

By telecon dated August 4, 1985, the licensee provided information relevant to the "special circumstances" finding required by revised 10 CFR 50.12(a) (See 50 Fed. Reg. 50764). The licensee stated that existing and proposed fire protection features at H. B. Robinson accomplish the underlying purpose of the rule.

Implementing additional modifications to provide additional suppression systems, detection systems, and fire barriers would require the expenditure of engineering and construction resources as well as the associated capital costs which would represent an unwarranted burden on the licensee's resources. The licensee stated that the costs to be incurred are as follows:

- IIl.G.2.f Engineering procurement and installation of radiant energy heat shields.

The licensee stated that these costs are significantly in excess of those required to meet the underlying purpose of the rule. The staff concludes that "special circumstances" exist for the licensee's requested exemptions in that application of the regulation in these particular circumstances is not necessary to achieve the underlying purposes of Appendix R to CFP Part 50. See 10 CFR 50.12(a) (2) (ii).

The technical requirements of Section IIl.G.2.f are not met in the containment area because certain alternate shutdown related instrument cables are not protected by a radiant energy heat shield and would not be free of fire damage after being involved in a fire.

Fire loading in containment includes 960 pounds of charcoal (2000 Btulft2) and 200 gallons of lube oil per RCP (20,000 Btulft2/Pump Bay). A fire severity of less than 20 minutes would be associated with the preceding fire loadings. By letter dated June 13, 1984 CP&L submitted results of a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire test which confirmed that ASTM E-1 19 criteria were met or exceeded.

Also, there is a concern that the heat produced in a fire would cause cable trays to collapse and impact the cable and cause its failure. The Rockbestos cables will be in conduits and supported on the missile wall on unistrut type supports which keep the cable/conduit close to the wall. Therefore, falling debris will not pull it down and because of the low fire loading, insufficient heat will be generated to cause support failure. The cable will be used at 24 VDC and have been tested at 11 OVAC, 480 VAC and 960VAC. Therefore, there is no concern about its use at high voltage.

Because the fire-rated cable would be damaged by a fire, there were concerns that this damage would affect the performance of the shutdown functions for a time period that is significantly longer than the time HBRSEP LAR Rev 0 Page K-4

Duke Energy Attachment K - Existing Licensing Action Transition period for which the function is required. The proposed use of this cable is to provide a radiant energy heat shield for use inside containment. The cables were subjected to an ASTM E-1 19 fire test, and the circuit integrity was maintained and-kept functional for a period of 93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br />. Rockbestos cables were previously evaluated and accepted by the NRC for use where a 1-hour fire-rated barrier was required by Appendix R.

This acceptance was granted for TMI via an NRC letter dated April 19, 1985. The intent of a radiant energy heat shield for use inside containment was to offer a lesser level of passive fire protection than a 1-hour barrier. This was in recognition of the fact that the containment fire hazards tended to be low and containment was not susceptible to the degree of transient fires expected to occur outside of containment. Therefore. it is concluded that the Rockbestos cable is quite conservative for the heat shield application, given minor fire damage that would be expected to occur inside containment.

For the distributed fire load in this area, it would be difficult to achieve a real fire that would result in temperatures approaching the ASTM E-1 19 time-temperature curve over a large portion of the fire area.

Prompt action by the fire brigade would further reduce the time-temperature curve. The hose stream tests with repeated application of hose stream forces have resolved this concern.

There was a concern that thermal expansion forces, and post-fire mechanical forces due to fire fighting and recovery operations, were not simulated. There was also a concern that "wet short" conditions were not simulated, in that cables in cable trays may be immersed in water for a significant time. The installation proposed by the CP&L is for conduits, and hose streams would not disrupt the cables. These cables, being in conduit, would not be immersed in water. These two concerns are resolved.

Based on the above evaluation, the staff concludes that the use of Rockbestos fire rated cables in lieu of a radiant energy heat shield inside containment provides a level of fire protection equivalent to the technical requirements of Section IIl.G.2.f of Appendix R.

Licensing 9/17/86 Date Transitioned E]

Reference Document Doc Detail NLS-84-242, Fire Protection - Appendix R, One Hour Rated Fire Barriers, 6/13/84 NLS-84-327, Appendix R Exemption Request, 8/17/84 NLS-86-570, Exemption from Certain Requirements of 10 CFR Part 50, Appendix R, Sections III.G.2 and III.G.3 - H. B. Robinson Steam Electric Plant, Unit No. 2, 9/17/86 Licensing Actions Licensing Area G3 Exemption from the requirements of Section III.G.2 of Appendix R to 10 CFR Part 50 Action Licensing Subsection III.G.2 specifies that one train of cables and equipment necessary to achieve and maintain hot Basis shutdown be maintained free of fire damage by one of the following means:

a. Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier;

b. Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal HBRSEP LAR Rev 0 Page K-5

Duke Energy Attachment K - Existing Licensing Action Transition 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; or

c. Enclosure of cables and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.

Service Water Pump Area The licensee requests an exemption from Section III.G.2 of Appendix R to the extent that it requires 1-hour fire rated barriers or 20 feet of separation free of intervening combustibles to separate redundant divisions and an automatic fire suppression system.

The service water pump area is located in the Intake Structure. This structure is formed by metal walls, concrete floor and an open roof. The area contains four service water pumps. One pump is needed for safe shutdown. The pumps are aligned with 2 to 4 feet separation between pumps. The separation between A and D pumps is approximately 13 feet. Control and power cables to the pumps are routed in conduit through the concrete floor and terminate directly into the motor end of the pumps.

The in-situ combustible loading is 6 gallons of lubricating oil contained in each of the service water pumps.

The fire protection in the area consists of manual hose stations and portable fire extinguishers. The licensee justifies this alternative on the following:

a) Manual fire fighting capability, b) Television Camera Surveillance of the area in lieu of fire detection, and c) An analytical model employed to show that the magnitude of an exposure fire needed to damage redundant components is significantly higher than reasonably expected.

The NRC staffs evaluation included the following considerations:

This area does not comply with Section III.G because it does not have an automatic suppression system and twenty feet of separation free of intervening combustibles. There is no alternate shutdown capability independent of this area and there is no automatic fire detection system. This area is under continuous television surveillance by security personnel. The in-situ combustible load is light. The only cables in the room are two short sections per pump that rise from the floor near each pump and terminate at the pump.

The lubricating oil is contained in the pump and there are no hot surfaces in the area.

The licensee has conducted an analysis to determine quantity of fuel, spilled on the floor of the area, that is needed to create a fire and corresponding heat flux of enough severity to cause cable damage. The analysis indicates that 17 gallons of acetone in an 8-foot-diameter pool, is needed to effect damage. Only administrative controls are available to prevent the accumulation of transient materials in individual plant areas. With the low combustible loading, and continuous surveillance in this area, there is reasonable assurance that a fire would be detected promptly and could be extinguished manually.

Based on our evaluation, the level of existing protection for this area does provide a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R and, therefore, the exemption for the Service Water Pump Area is granted.

Licensing 11125183 Transitioned E]

Date Reference Document Doc Detail APR2782, Fire Protection - Appendix R Exemption Requests, 4/27/82 HBRSEP LAR Rev 0 Page K-6

Duke Energy Attachment K - Existing Licensing Action Transition MAR1682, Conformance to the Requirements of 10CFR50, Appendix R, Section III.G and Response to Generic Letter 81-12, 3/16/82 NLU-83-777, Safety Evaluation Related to Exemptions from 10CFR50 Appendix R, 11/25/83 Licensing Actions Licensing Area H Exemption from the requirements of Section III.G.2 of Appendix R to 10 CFR Part 50 pertaining to Action the requirement for 3-hour rated barriers be installed to separate redundant trains Licensing Subsection II I.G.2 specifies that one train of cables and equipment necessary to achieve and maintain hot Basis shutdown be maintained free of fire damage. The NRC concluded that additional modifications would not enhance fire protection safety above that provided by existing and proposed alternatives for the facility and therefore exemptions are granted for Subsection III.G in the RHR Pit- Fire Zone 27.

On November 19, 1980, the Commission published a revised Section 10 CFR 50.48 and a new Appendix R to 10 CFR 50 regarding fire protection features of nuclear power plants (45 FR 76602). The revised Section 50.48 and Appendix R became effective on February 17, 1981.Section III of Appendix R contains fifteen subsections, lettered A through 0, each of which specifies requirements for a particular aspect of the fire protection features at a nuclear power plant. Subsection III.G.2 requires that one train of cables and equipment necessary to achieve and maintain safe shutdown be maintained free of fire damage by one of the following means:

a. Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier;

b. 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. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; or

c. Enclosure of cables and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area.

By letters dated January 9, 1981, March 1, 1982, and April 27, 1982, the licensee requested an exemption from the requirements of Subsection III.G.2. The licensee requested an exemption from III.G.2 of Appendix R to the extent that it requires 3-hour fire rated barriers be installed to separate redundant trains.

Fire Zone 27 is located west of the Auxiliary Building at elevation 203 feet. The area is separated from other plant areas by concrete walls. Entrance into the area is through a hatch and down a 25-foot ladder. The ceiling height in the area is 27 feet 6 in. The fire protection in the area consists of smoke and heat detectors, standpipe hose stations and portable fire extinguishers.

Fire Zone 27 contains two residual heat removal (RHR) pumps and associated piping. No equipment or circuit within this zone is required for hot shutdown, but one train of the RHR system is necessary to achieve and maintain cold shutdown. Each RHR pump is mounted on a concrete pedestal approximately 4ft. high with the top of the pump about 10 ft. above floor elevation. The redundant pumps are separated by a 22-ft-HBRSEP LAR Rev 0 Page K-7

Duke Energy Attachment K - Existing Licensing Action Transition high concrete barrier which completely bisects the RHR pit into two individual pump bays. Each pump bay has a sump approximately 3 ft. x 3 ft. x 6 in. deep with an installed sump pump. The sumps are adjacent to each other and separated by the same barrier which divides the zone. A hole approximately 4 in. in diameter joins the sumps so each sump pump can serve as a backup to the other.

The licensee indicates that a fire in the RHR pit would cause damage to both trains of the RHR control and power cables; however, the licensee also indicates that the cables could be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the time allowed by Section III.G.I.

The combustible in Fire Zone 27 is lubricating oil contained in the RHR pumps. Each pump contains 8 gallons of oil. The oil comprises a fuel load of 6500 Btu/sq. ft. which if totally consumed, would correspond to a fire severity of about 5 minutes on the ASTM E-1 19 standard time temperature curve.

The licensee justifies this exemption based on the following:

a) Access to the area is limited.

b) The in-situ combustible loading is light c) Smoke and heat detection are provided.

d) A partial height (22 feet) concrete barrier separates the RHR pumps.

e) An analytical model was employed to show that the magnitude of an exposure fire needed to ignite the in-situ lube oil is significantly higher than reasonably expected.

f) Hot surfaces necessary to cause ignition of the lube oil do not exist in the area.

This area does not comply with Section III.G because the redundant RHR trains are not separated by 3-hour fire rated barriers, there is no automatic suppression system, and there is no alternate cold shutdown capability independent of the area.

The NRC staff has evaluated this area with the following considerations. This area is normally locked and the pumps are not running. Even with the pumps running, there are no hot surfaces in the area. The few cables for the pumps are in conduit. The only sigrificant in-situ combustible in the fire area is the pump motor lubricating oil. The probability of ignition of the oil is low because the lubricating oil has a high flashpoint (approximately 450 degrees) and sufficiently hot surfaces do not exist in this fire area to cause the ignition of the lube oil. If a fire occurred in the RHR pit, access into the pit for nominal fire fighting would be difficult due to the smoke and hot products of combustion that would vent through the hatch entrance. We anticipate that manual fire fighting activities would be conducted from hatch entrance rather than from inside the RHR pit. This may result in water damage to both trains of RHR. However. with proper fire fighting procedures along with the concrete wall separating the pumps, such damage could easily be prevented. In addition, because the RHR pumps are only used during cold shutdown, there are emergency procedures which could be used to maintain safe conditions in the unlikely event of a fire and fire fighting activities that affect both pumps.

Because the area is normally locked, a partial height wall separates the RHR pump, the area contains few combustibles, and the pumps are only needed for cold shutdown, an automatic suppression system is not necessary. The fire detectors should assure prompt detection of a fire should it occur. This arrangement would provide reasonable assurance that a fire would not damage both RHR pumps and the damage to cables would be limited so that it could be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Based on the above evaluation, the level of existing fire protection for this area does provide a level of fire protection equivalent to the technical requirements of Section III.G of Appendix Rand, therefore, the licensee's exemption request is granted for this area.

Licensin 112583 Transitioned []

HBRSEP LAR Rev 0 Page K-8

Duke Energy Attachment K - Existing Licensing Action Transition Date Reference Document Doc Detail APR2782, Fire Protection - Appendix R Exemption Requests, 4/27182 MAR1 682, Conformance to the Requirements of 10CFR50, Appendix R, Section Ill.G and Response to Generic Letter 81-12, 3/16/82 NLU-83-777, Safety Evaluation Related to Exemptions from 10CFR50 Appendix R, 11/25/83 Licensing Actions Licensina Exemption from the Requirements of Section 111.0 of Appendix R to 10 CFR Part 50 Action Licensino Exemption from Section 111.0 of Appendix R is granted by the NRC to the extent that a reactor coolant pump Basis lube oil collection system is not provided. In lieu of installing such a system, fixed fire suppression is maintained and additional detection and dikes were installed in the pump bays. Also, the Containment Spray system serves as a backup fire suppression system with Sodium Hydroxide isolated.

By letter dated November 16, 1980, the licensee informed the staff that installation of a fixed fire suppression system had been completed in each reactor coolant pump bay. The system was based on proposed NRC requirements set forth in Section Ill.P of a proposed Appendix R notice in the Federal Register on May 29, 1980. Additional information was provided by letter dated November 26, 1980. By letter dated January 19, 1981, the licensee requested an exemption from the requirements of Section 111.0 of Appendix R to 10 CFR 50. The request was supplemented by letters dated January 7, 1983, July 30, 1982, June 7, 1983, June 29, 1983 and October 5, 1983. The June 7, 1983, and October 5, 1983 letters proposed modifications for additional fire suppression as alternatives. This does not meet the technical requirements of Appendix R because oil collection systems for the reactor coolant pumps are not provided. The acceptability of the alternative measures are discussed below.

The containment contains three reactor coolant pumps (A, B and C). These are located in bays (A, B and C). These bays also contain safety related cabling for the reactor coolant loop instrumentation. Bays A and B share a common ceiling; Bay C is isolated from Bays A & B to some extent. The bays are covered by removable concrete blocks. These blocks will cause the plume from an unmitigated fire to be diverted through the steam generator area. This area contains safety related steam flow instrumentation sensing lines.

Oil spilled in Bay A, will be confined to Bay A; however, oil spilled in Bays Band C can flow to adjacent areas. The foundation for the reactor coolant pumps is at the 237.000' level. The foundation for the steam generators is at the 238.33' level. The reactor coolant pump is located between the pressurized portion of the oil system and the steam generator supports, and serves to shield the steam generator supports in the event of an oil system rupture.

The major combustible in each bay is the 200 gallons of oil in each reactor coolant pump.

The existing fire detection system in each reactor coolant pump bay is a two-zone detection system. One zone consists of a single infrared flame detector; the other zone consists of a 325°F fixed-temperature heat detector. Activation of one zone of detection sends an alarm to the control room; activation of the second HBRSEP LAR Rev 0 Page K-9

Duke Energy Attachment K - Existing Licensing Action Transition zone of detection alarms in the control room and also opens the preaction water deluge valve to the bay.

Both detectors are wall mounted.

The existing fire suppression system for each bay, is a preaction sprinkler system. Each bay has its own deluge valve, supply header, and a ring header that encircles the reactor coolant pumps at elevation 239 feet 4 inches. Each of the five risers off the ring header have three 220°F closed head side wall sprinklers at approximately 240 feet, 245 feet and 252 feet elevations. These systems are designed to meet the minimum residual pressure and flow requirements of NFPA-Std-15.

The suppression system ring header piping in Bay A is designed to withstand an SSE, while Bays B and C are designed such that a seismic event would not impact safety related equipment due to suppression system rupture. The risers are restrained to withstand the nozzle reaction forces. These forces are greater than those anticipated from a seismic event.

The existing containment spray system would be used as an emergency back-up to the bay suppression system if necessary to cool the operating level and containment annulus outside of the RCP bays.

By letter dated June 7, 1983, the licensee proposed to:

(1) Provide additional ceiling mounted heat detectors to meet the spacing and location requirements of NFPA-STD-72E, ""Standard on Automatic Fire Detectors.""

(2) Replace existing closed head sprinklers with special open water spray nozzles and manual actuation from the control room.

(3) Construct 6 inch dikes at the 231 feet elevation in Bay B and Bay C.

(4) Revise operating procedures for the containment spray system to allow its operation as a back up fire suppression system with the sodium hydroxide valves out.

By letter dated October 5, 1983, the licensee committed to maintain an automatically actuated closed-head preaction system in lieu of a manually actuated open-head system.

We have evaluated the fire protection for the reactor coolant pump lube oil system and conclude that the effects of a fire in an RCP Bay will not prevent safe shutdown capability. There are no components within the RCP Bay that are required for safe shutdown. The effects of any fire within an RCP Bay will be prevented from affecting the safe shutdown equipment outside the RCP Bay by the suppression system inside the RCP Bay and the Containment Spray System outside the Bay.

It is the staffs conclusion that: 1) installation of a reactor coolant pump oil collection system in this facility would not Significantly enhance fire safety, and 2) the existing fire protection system in the Reactor Coolant Pump Bays with the addition of the proposed modifications provides an acceptable level of safety to that achieved by compliance with the requirements of Section 111.0 of Appendix R to 10 CFR 50. Therefore, the licensee's request for an exemption should be granted.

Licensing 317185 DateTransitioned i]

Date Reference Document Doc Detail Jan783, Requested Information To Support Exemption From The Requirements Of 10CFR50, Appendix R Section 111.0, 117/83 July3082, Appendix R, Section 111.0, Request' For Exemption Reactor Coolant-Pump Lubricating Oil Collection System, 7/30/82 LAP-83-210, Additional Information Conceming Pending Exemption Requests -

Appendix R, 6/7/83 HBRSEP LAR Rev 0 Page K-10

Duke Energy Attachment K - Existing Licensing Action Transition LAP-83-261, Clarifying Information - Appendix R Modifications, 6/29/83 LAP-83-422, Supplemental Information Concerning Pending Exemption Requests - Appendix R, 10/5/83 NLS-85-146, Exemption - Appendix R to 10CFR Part 50, Item 111.0, H. B.

Robinson Steam Electric Plant Unit No. 2, 3/7/85 NLS-85-176, RCP Oil Collection Exemption, 3/7/85 NO-80-1654, Fire Protection Status, 11/6/80 NO-80-1752, Fire Protection for Reactor Coolant Pump Bays, 11/26/80 NO-81-111, Fire Protection for Reactor Coolant Pump Bays, 1/19/81 NO-81-448, Petition of Carolina Power & Light Company for Exemptions from Certain Requirements of 10 CFR 50.48 and Appendix R to 10 CFR Part 50, 3/11/81 Licensing Actions Licensing NRC Acceptance Fire Hydrants Action Licensing The NRC accepted the spacing of the fire hydrants installed on the site. The basis was that the Basis configuration of the fire hydrants provides adequate coverage.

Licensing 2/28/78 Date DateTransitioned El Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 Licensing Actions Licensing NRC Acceptance of Building Separation Action Licensing The NRC accepted the adequacy of building separation at the site. The barriers were considered to be Basis adequate for the hazards. Fire areas included are A5, A10, A13, A14, A15, A16, A17, A18, A19, G1, G4, and G7.

Licensing 2/28/78 DateTransitioned Reference Document Doc Detail NG-77-704, Fire Protection Program Review, 6/23/77 Question 18 NLU-78-71, License Amendment 31, 2/28/78 Section 4.11, 4.14 HBRSEP LAR Rev 0 Page K-1 1

Duke Energy Attachment K - Existing Licensing Action Transition Licensing Actions Licensing NRC Acceptance of Class II Hose Stations Action Licensing The NFPA 805 code requirement is for Class IlIIhose stations, which consist of both a 1-1/2" and 2-1/2" Basis hose connection. The site has Class II stations, which only have a 1-1/2" hose connection. This configuration was approved by the NRC.

Licensina 2/28/78 DataTransitioned []

Date Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 Licensing Actions Licensing NRC Acceptance of Compatible Thread Connections Action Licensing The NRC approved the types of compatible hose thread connections used at the site.

Basis Licensing 2/28/78 Transitioned []

Date Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 Licensing Actions Licensing NRC Acceptance of Fire Barriers Action Licensing The NRC approved the approach used for fire barriers at the plant. The barriers were considered to be Basis adequate for the hazards. Fire areas included are A5, A10, A13, A14, A15, A16, A17, A18, A19, G1, G4, and G7.

Licensing 2/28/78 Transitioned F1 Date Reference Document Doc Detail NG-77-704, Fire Protection Program Review, 6/23/77 Question 18 NLU-78-71, License Amendment 31, 2/28/78 Section 4.11, 4.14 HBRSEP LAR Rev 0 Page K-12

Duke Energy Attachment K - Existing Licensing Action Transition Licensing Actions Licensing NRC Acceptance of Fire Pump Separation Action Licensing In the 1977 submittal, CP&L provided a description of the physical separation of the fire pumps, including Basis the fact that the pumps were not separated by three hour rated fire walls. The NRC accepted the configuration in the 1978 license submittal.

Licensing 2/28/78 Transitioned Date Reference Document Doc Detail NG-77-704, Fire Protection Program Review, 6/23/77 Question 15 NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.1 Licensing Actions Licensing NRC Acceptance of Internal Conduit Fire Seals Action Licensing Section (b): The internal conduit seals utilized at HBRSEP are based on fire test conducted by a consortium Basis led by Wisconsin Electric. The test report was submitted to the NRC who issued a Technical Evaluation Report on the program.

Licensing 5/12/89 Transitione,,

PAq Reference Document Doc Detail 8911030114, Review of Draft Safety Evaluation Of Conduit Fire Seal Topical Report for Propietary Content, 10/23/89 CTL# CRE093-4324, Conduit Fire Test of One Hundred One Electrical Conduit Penetrations, 6/1/87 Licensing Actions Licensing NRC Acceptance of Non-Seismic Standpipes Action Licensing In the Federal Register Notice that promulgated 10CFR50.48(c), the NRC stated that plants that were Basis originally reviewed under Appendix A to BTP APCSB 9.5-1 were exempted from the requirements of section 3.6.4 of NFPA 805.

Licensing 6/16/04 Transitioned El Date...

Reference Document Doc Detail HBRSEP LAR Rev 0 Page K-1 3

Duke Energy Attachment K - Existing Licensing Action Transition 69 FR 33536, Federal Register Notice, Voluntary Fire Protection Requirements for Light Water Reactors, 6/16/04 Licensing Actions Licensing NRC Acceptance of Site Fire Water Supply Action Licensing The site complies with Exception No. 1. The NRC approved the existing installation.

Basis Licensing 2/28/78 Date Tranitoned El Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Licensing Actions Licensing NRC Acceptance of Water Supply Pump Connection Requirements Action Licensing The NRC accepted the fire pump connection to the fire main.

Basis Licensing 2/28/78 Date Transitioned El Date Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 Licensing Actions Licensing NRC Approval of Fire Suppression Affects Action Licensing The NRC approved the impact of fire water system rupture in the auxiliary building hallway on Elevation Basis 226. The basis was the installation of a water shield to protect MCC No. 5.

Licensing 12/8/80 DateTransitioned Reference Document Doc Detail NLU-80-623, NRC Safety Evaluation Report, 12/8/80 Section 3.2.7 NO-80-896, Fire Protection Program RAI Responses, 6/12/80 Item 3.2.7 HBRSEP LAR Rev 0 Page K-14

Duke Energy Attachment K - Existing Licensing Action Transition Licensing Actions Licensing NRC Approval of non-IEEE 383 cables Action Licensing The NRC approved the use of non-IEEE 383 rated cables. The basis for this would be that cables which did Basis not meet the flame test requirements would be coated with flame retardant coating. Fire areas included are A3, A5, A6, A9, A13, A14, Al5, Al6, A17, Al8, A19, C, D, and E.

Licensing 7/28/78 Date Transitioned El Date Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.8 Licensing Actions Licensing NRC Approval of Penetration Seal Rating Action Licensing The NRC approved penetration seals in the following rooms as having a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire rating. Fire areas Basis included are A3, A5, A6, A8, A9, A10, All1, A12, A13, A14, A15, A16, A17, A18, A19, C, D, and E.

The basis for all areas except the North Cable Vault (fire area D) was the low combustible loading. The North Cable Vault took credit for the fire detection, automatic C02 system, and manual fire fighting capability.

Licensing 11/25183 Transitloned El Date Reference Document Doc Detail NLU-83-777, Safety Evaluation Related to Exemptions from IOCFR50 Appendix Enclosure Page 5 R, 11/25/83 Licensing Actions Licensing NRC Approval of Yard Main System Action Licensing The NRC approved the overall plant fire main layout in the 1978 SER.

Basis Licensina 2/28178 Transitioned El Date HBRSEP LAR Rev 0 Page K-1 5

Duke Energy Attachment K - Existing Licensing Action Transition Reference Document Doc Detail NG-77-704, Fire Protection Program Review, 6123177 Question 15 NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 Licensing Actions Licensing NRC Approval of Yard Sectionalizing Valves Action Licensing The NRC approved the sectionallizing valves installed in the fire main in the 1978 SER. The basis was that Basis an isolation valve would be added at the connection of the Unit 1 loop with the Unit 2 loop, and separate headers would be installed for sprinkler systems in the reactor auxiliary building. In addition, vehicle barriers were provided around all hydrants and post indicator valves.

Licensing 2/28/78 Data Trensitioned El Date Reference Document Doc Detail NLU-78-71, License Amendment 31, 2/28/78 Section 4.3.1.3 HBRSEP LAR Rev 0 Page K-16

Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval L. NFPA 805 Chapter 3 Requirements for Approval 10 CFR 50.48(c)(2)(vii) 13 Pages Attached Page Li HBRSEPLARRevO HBRSEP LAR Rev 0 Page L-1

Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 1 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."

HBRSEP has wiring above suspended ceilings that may not comply with the requirements of this code section.

Suspended ceilings are noncombustible and exist only in the Control Room (FZ 23),

Inside AO Office and old Turbine Building RCA Entrance (FZ 25A). Combustibles in concealed spaces are minimal.

The three areas currently with suspended ceilings inside the NFPA 805 defined power block are in the Control Room (FZ 23), Inside AO Office and old Turbine Building RCA Entrance (FZ 25A). The Inside AO Office and old Turbine Building RCA Entrance (FZ 25A) are not risk significant. Neither of the rooms nor the cables are safety-related.

Most electrical wiring above the Control Room partial suspended ceiling is in conduit except for short flexible connectors to lighting fixtures. There is one eight-foot length of eight-inch diameter UL approved flexible air duct with flame spread rating of 25 or less.

The quantity of cabling above the suspended ceilings in the Control Rooms is very low and results in limited combustible loading. In addition, the existing fire detection capability and/or the Control Room Operators who are continuously present in the area would identify the presence of smoke.

These areas are assumed to have wiring above the suspended ceilings including that needed for lighting, power, control, and video/communication/data. Power and control cables at HBRSEP are IEEE-383-1974 or equivalent. FAQ 06-0022 identified acceptable electrical cable construction tests. Plenum rated cable is tested to NFPA 262 and the FAQ concluded that the NFPA 262 test is equivalent to the IEEE-383-1974 test.

Therefore, IEEE cable is inherently equivalent to plenum rated cable and acceptable to be routed above suspended ceilings.

Video/communication/data cables that have been field routed above suspended ceilings are low voltage. Existing cables for video, communication, and networking may not be plenum rated, but are not generally susceptible to shorts that would result in a fire.

Basis for Request:

The basis for the approval request of this deviation is:

" Power and control cables comply with this section (plenum rated equivalent or armored).

" The wiring above ceilings in office areas does not pose a hazard:

" Low voltage is not susceptible to shorts causing a fire.

" Power and control cables are protected (plenum rated equivalent or armored) per this code section.

HBRSEP LAR Rev 0 Page L-2

Duke Ener_(v Attachment L - NFPA 805 Chai)ter 3 Reauirements for Agmroval

• By eliminating cables with the potential shorts, this eliminates ignition sources and, therefore, the jacketing of cable is not relevant.

• No equipment important to nuclear safety is located in the vicinity of these cables.

• New or replacement cables are plenum rated and constructed similar to or superior to the original cable and meet the requirements of IEEE-383-1974.

Existing fleet procedures will be used to ensure that changes moving forward are considered for NFPA 805 impacts. (FIR-NGGC-0010)

Acceptance Criteria Evaluation:

Nuclear Safety and Radiological Release Performance Criteria:

The location of wiring above suspended ceilings does not affect nuclear safety. Power and control cables comply with this section. Other wiring, while it may not be in armored cable, in metallic conduit, 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 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 cables or locations of suspended ceilings. The location of cables does not change the radiological release evaluation performed that potentially contaminated water is contained and smoke monitored. The cables do not add additional radiological materials to the area or challenge system boundaries that contain such.

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Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval Safety Margin and Defense-in-Depth:

Power and control cables meet the requirements of this requirement. The use of these materials has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these 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 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 current cable installation procedures documenting the requirements of NFPA 805 Section 3.3.5.1. The control room is a continuously manned area of the plant. 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 result in compromising automatic fire suppression functions, manual fire suppression functions, fire protection for systems and structures, or post-fire safe shutdown capability.

==

Conclusion:==

HBRSEP determined that the performance based approach satisfies the following criteria:

" Satisfies the performance goals performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release

• Defense in Depth

• Safety Margin Page L-4 HBRSEPLARRevO HBIRSEP LAR Rev 0 Page L-4

Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval Approval Request 2 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 electricalraceways. 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 HBRSEP per HBR2-0B060 Sht D6 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 noncombustible 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.

Basis for Request:

" 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.

Acceptance Criteria Evaluation:

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.

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Duke Energy Aftachment L - NFPA 806 Chapter 3 Requirements for Approval 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 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:==

HBRSEP determined that the performance based approach satisfies the following criteria"

" Satisfies the performance goals performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release

" Defense in Depth

" Safety Margin Page L.8 HBRSEPLARRevO HBRSEP LAR Rev 0 Page L-6

Duke Energy Attachment L - NFPA 806 Chapter 3 Requirements for Approval Approval Request 3 NFPA 805 Section 3.5.16 NFPA 805 Section 3.5.16 states:

"The fire protection water supply system shall be dedicated for fire protection use only.

Exception No. 1: Fireprotection water supply systems shall be permitted to be used to provide backup to nuclearsafety systems, provided the fire protection water supply systems are designed and maintainedto deiver the combined fire and nuclearsafety 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 durationas determinedin this section.

The review of plant flow diagrams show no hard connections to other plant systems, besides those for fire protection use. It should be noted that although there are no hard pipe connections to other plant systems, there are procedures that utilize the fire protection water supply. They are as follows:

" AOP-014 - Loss of CCW

" AOP-022 - Loss of Service Water

• EDMG-001 - Extreme Damage Event Early Actions

• EDMG-002 - Refueling Water Storage Tank (RWST)

• EDMG-003 - Condensate Storage Tank (CST)

" EDMG-005 - Containment Vessel (CV)

" EDMG-01 1 - Spent Fuel Pit Casualty

" EDMG-012 - Core Cooling Using Alternate Water Source

• EDMG-013 - Airborne Release Scrubbing

" SAM Inject into the Steam Generator

• SAM Inject into the RCS

• SAM Inject into Containment

" SAM Control Containment Conditions

" SAM Flood Containment The use of the fire protection water for these non-fire protection system water demands would have no adverse impact on the ability of the fire protection system to provide required flow and pressure. OMM-002, Section 8.15, details restrictions and allowances for use of the fire protection water supply system at HBRSEP.

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Duke Energly Attachment L - NFPA 805 Chapter 3 Requirements for Approval Basis for Request:

The use of the fire protection water for these non-fire protection system water demands would have no adverse impact on the ability of the fire protection system to provide required flow and pressure. This is based on how fire water usage is restricted (CR 99-01247), in the following ways:

1. Fire service related activities (emergency, testing and training).

2. When the use of fire water is specifically called out in approved plant procedures (i.e., AOPs).

3. During plant emergencies when fire water is needed to protect safety related equipment.

4. When usage is deemed necessary AND sufficient justification is provided to show that the use of the fire water system for the proposed activity does not cause the fire water system to be in a condition outside of its design basis (i.e., the quantity of water needed for the proposed activity does not drop supply and pressure below that required/defined in UFSAR Section 9.5.1).

Permission shall have the approval of the Shift Manager (CR 96-00729 and CR 96-00730).

The water supply system is capable of maintaining the pressure in the main plant loop at 70 psi or higher with the largest deluge system in operation and with the system supplying an additional 1000 gpm to hoses.

Acceptance Criteria Evaluation:

Nuclear Safety and Radiological Release Performance Criteria:

The use of fire protection water for non-fire protection plant evolutions is an occurrence that requires Shift Manager review and concurrence. The flow limitations to those non-fire protection functions ensure that there is no impact in the ability of the automatic suppression systems to perform Therefore, there is no impact on the nuclear safety performance criteria.

The use of fire protection water for plant evolutions other than fire protection has no impact on the radiological release performance criteria. The radiological release performance criteria is satisfied based on the determination of limiting radioactive release (Attachment E), which is not affected by impacts on the fire protection system due it's use for non-fire protection purposes.

Safety Margin and Defense-in-Depth:

The use of the fire water system, including the use of hydrants and hose, for non-fire protection uses does not impact fire protection defense-in-depth. The fire pumps have the excess capacity to supply the demands of the fire protection system as well as the non-fire protection uses identified above. This does not compromise automatic or manual fire suppression functions, fire suppression for systems and structures, or the nuclear safety capability assessment. Since both the automatic and manual fire suppression functions are maintained, defense-in-depth is maintained.

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Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval The methods, input parameters, and acceptance criteria used in this analysis were reviewed and found to be in accordance with NFPA 805 Chapter 3. The methods, input parameters, and acceptance criteria used to calculate flow requirements for the automatic and manual suppression systems were not altered. Therefore, the safety margin inherent in the analysis for the fire event has been preserved.

==

Conclusion:==

HBRSEP determined that the performance based approach satisfies the following criteria:

• Satisfies the performance goals performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release

" Defense in Depth

" Safety Margin Page L-9 HBRSEP HBRSEP LAR Rev 0 LAR Rev 0 Page L-9

.Duke Energqy Aftachment L - NFPA 805 ChaDter 3 Reaulrements for Aimroval Approval Request 4 NFPA 805 Section 3.2.3(1)

In accordance with 10 CFR 50.48(c)(2)(vii), "Performance-based methods," the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard.

In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied.

In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

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 nuclear safety capability).

Duke Energy, HBRSEP requests formal approval of performance-based exception to the requirements in Chapter 3 of NFPA 805 as follows:

NFPA 805, Section 3.2.3(1)

"Proceduresshall be established for implementation of the fire protection program.

In addition to proceduresthat could be requiredby other sections of the standard, the proceduresto accomplish the following shall be established:

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

Duke Energy, HBRSEP requests the ability to utilize performance-based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. Performance-based inspection, testing, and maintenance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report TR-1 006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection", Final Report, July 2003.

Basis for Request:

NFPA 805 Section 2.6, "Monitoring," requires that "A monitoringprogram shall be established to ensure that the availabilityand reliabilityof the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria.

Monitoring shall ensure that the assumptions in the engineeringanalysis remain valid."

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Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval NFPA 805 Section 2.6.1, 'Availability, Reliability, and PerformanceLevels, "requires that "Acceptablelevels of availability,reliability, and performance shall be established."

NFPA 805 Section 2.6.2, "MonitoringAvailability, Reliability, and Performance,"

requires that "Methods to monitor availability,reliability, and performance shall be established. The methods shall consider the plant operatingexperience and industry operating experience."

The scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the fire protection program have been established based on the previously approved Technical Specifications / License Controlled Documents and appropriate NFPA codes and standard. This request does not involve the use of the EPRI Technical Report TR-1 006756 to establish the scope of those activities as that is determined by the required systems review identified in Attachment C This request is specific to the use of EPRI Technical Report TR-1 006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program. As stated in EPRI Technical Report TR-1006756 Section 10.1, "The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability." This goal is consistent with the stated requirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1 006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained.

The target tests, inspections, and maintenance will be those activities for the NFPA 805 required fire protection systems and features. The reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The failure criterion will be established based on the required fire protection systems and features credited functions and will ensure those functions are maintained.

Data collection and analysis will follow the EPRI Technical Report TR-1006756 document guidance. The failure probability will be determined based on EPRI Technical Report TR-1006756 guidance and a 95% confidence level will be utilized. The performance monitoring will be performed in conjunction with the Monitoring Program required by NFPA 805 Section 2.6 and it will ensure site specific operating experience is considered in the monitoring process. The following is a flow chart that identifies the basic process that will be utilized.

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Duke Energy Attachment L - NFPA 806 Chapter 3 Requirements for Approval Duke Energy Affachment L NFPA 805 Chapter 3 Requirements for Approval Program Framework Identify Target Tests and Inspections Establish Reliability and Frequency Goals Set Failure Criteria Assess Licensing Impact and Other Constraints I

Data Collection and Evaluation Establish Data Collection Guidelines Collect Required Surveillance Data Assemble Data in Spreadsheet or Database I

Analyze Data to Identify Failures Reliability and Uncertainty Analysis Compute Failure Probabilities Compute Uncertainty Limits I

Confirm That Reliability Supports Target Frequency Program Implementation Modify Program Documents Revise Surveillance Procedures Conduct Ongoing Performance Monitoring Refine and Modify Frequencies as Appropriate EPRI TR-1 006756 - Figure 10-1 Flowchart for Performance-Based Surveillance Program HBRSEP LAR Rev 0 Page L-12

Duke Energy Attachment L - NFPA 806 Chapter 3 Requirements for Approval Duke Energy, HBRSEP does not intend to revise any fire protection surveillance, test or inspection frequencies until after transitioning to NFPA 805. Existing fire protection surveillance, test and inspection will remain consistent with applicable station, Insurer, and NFPA Code requirements. HBRSEP's intent is to obtain approval via the NFPA 805 Safety Evaluation to use EPRI Technical Report TR1 006756 guideline in the future as opportunities arise. Duke Energy, HBRSEP reserves the ability to evaluate fire protection features with the intent of using the EPRI performance-based methods to provide evidence of equipment performance beyond that achievable under traditional prescriptive maintenance practices to ensure optimal use of resources while maintaining reliability.

Nuclear Safety and Radiological Release Performance Criteria:

Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to Nuclear Safety Performance Criteria by the use of the performance-based methods in EPRI Technical Report TR-1006756.

The radiological release performance criteria are satisfied based on the determination of limiting radioactive release. Fire Protection Systems and Features may be credited as part of that evaluation. Use of performance-based test frequencies established per the EPRI Technical Report TR-1 006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited to meet the Radioactive Release performance criteria. Therefore, there is no adverse impact to Radioactive Release performance criteria.

Safety Margin and Defense-in-Depth:

Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited in the Fire Risk Evaluation safety margin discussions. In addition, the use of these methods in no way invalidates the inherent safety margins contained in the codes and standards used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analysis has been preserved.

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Duke Energy Aftachment L - NFPA 806 Chapter 3 Requirements for Approval The three echelons of defense-in-depth described in NFPA 805 Section 1.2 are

1) to prevent fires from starting (combustible/hot work controls),

2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and

3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions).

Echelon 1 is not affected by the use of the EPRI Technical Report TR-1 006756 methods. Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features credited for defense-in-depth are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2 and 3 for defense-in-depth.

==

Conclusion:==

NRC approval is requested for use of the performance-based methods contained in the Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide", Final Report, July 2003 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. As described above, this approach is considered acceptable because it:

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).

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Duke Energy Attachment M - License Condition Chancies Duke Energy Attachment M License Condition Changes M. License Condition Changes 3 Pages Attached Page M-1 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page M-1

Duke Enerav Attachment M - License Condition Channes Replace the current HBRSEP fire protection license condition 3.E with the standard license condition from Regulatory Guide 1.205, modified as shown below. No other license conditions need to be superseded or revised.

HBRSEP 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 HBRSEP Operating License DPR-23, by HBRSEP licensing staff and Duke Energy fire protection staff. The review was performed by reading the Operating License and performing electronic searches. Outstanding License Amendment Requests that have been submitted to the NRC were also reviewed for potential impact on the license conditions.

Supersede the existing license condition 3.E, in its entirety, as shown below:

E. Fire Protection Program Carolina Power & Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Report dated February 28, 1978, and supplements thereto.

Carolina Power & Light Company may make changes to the approved Fire Protection Program without prior approval 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 HBRSEP's understanding that implicit in the superseding of this license condition, all prior fire protection program SEs and commitments have been superseded in their entirety by the revised license condition.

The proposed license condition follows:

Duke Energy 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 license amendment request dated September 25, 2012, and as approved in the safety evaluation report 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 below are satisfied.

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Duke Energy Attachment M - License Condition Changes Risk-Informed Changes that May Be Made Without Prior NRC ADDroval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data 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 change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 E-7/yr for CDF and less than 1E-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 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 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 HBRSEP LAR Rev 0 Page M-3

Duke Energy Attachment M- License Condition Changes 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).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

(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 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]

(3) The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.

Duke Energy may perform change evaluations for deviations from the codes, standards, and listings referenced in NFPA 805, without a 10 CFR 50.90 submittal, as long as the specific requirement for the feature is not included in NFPA 805 Chapter 3, and the NFPA 805 change process is used.

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Duke Energy Attachment N - Technical Specification Changes N. Technical Specification Changes 2 Pages Attached Page N-I HBRSEP LAR Rev 0 LAR Rev 0 Page N-1

Duke Energy Affachment N - Technical Specification Changes Technical Specification (TS) 5.4.1.d will be deleted, existing TS 5.4.1.c will be changed, and existing TS 5.4.1 .e will be changed to TS 5.4.1 .d:

5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

d. Firc Protcction Programn implementation; and-.

No other TSs or Bases need to be revised.

Marked-up Technical Specifications are attached.

HBRSEP implemented the following process for determining that these are the only TSs 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 10 CFR 50.48(c).

2 A review was conducted of the HBRSEP TSs by Duke Energy fire protection staff and confirmed by HBRSEP regulatory affairs staff. The review was performed by reading the TSs and performing electronic searches. Outstanding LARs that have been submitted to the NRC were also reviewed for potential impact on the license conditions.

HBRSEP determined that these changes to the TSs are adequate for HBRSEP's adoption of the new fire protection licensing basis, for the following reasons.

E The requirement for establishing, implementing, and maintaining fire protection procedures is now contained in the regulation (10 CFR 50.48(a) and 10 CFR 50.48(c) NFPA 805 Chapter 3).

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Duke Energy Attachment N - Technical Specification Changes Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating procedures required to implement the commitments to NUREG-0737 and of NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Quality assurance for effluent and environmental monitoring: and

[Delete 4d , *.

P . ., ._ , *..

dUt. A.l Protectionpecified inSplmetfation; and d.e- All programs specified in Specification 5.5.

HBRSEP Unit No. 2 5.0-6 Amendment No. 176 RNP LAR Rev 0 Page N-3

Duke Energly Aftachment 0 - Orders and Exemption Duke Energy Attachment 0- Orders and Exemption

0. Orders and Exemptions I Page Attached Page 0-1 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page 0-1

Duke Energy Affachment 0 - Orders and Exemption Exemptions Rescind the following exemptions granted against 10 CFR 50, Appendix R dated 11/25/83 NLU-83-777 NRC Letter Regarding Exemption to 10 CFR 50 Appendix R Subsection III.G and III.M 10/25/84 NLU-84-687 NRC - Appendix R to 10CFR50, Items III.G.2 09/17/86 NLS-86-570 NRC Letter Regarding Exemption from Certain Requirements of 10 CFR Part 50, Appendix R, Sections III.G.2 and III.G.3 10/17/90 NRC-90-622 NRC Letter Regarding Exemption From Requirements of Section III.G.2 of Appendix R of 10 CFR Part 50 03/03/85 NLS-85-176 RCP Oil Collection Exemption Specific details regarding these exemptions are contained in Attachment K.

Orders No Orders need to be superseded or revised.

HBRSEP implemented the following process for making this determination:

A review was conducted of the HBRSEP docketed correspondence by HBRSEP licensing staff and the NFPA 805 Transition Team. The review was performed by reviewing the correspondence files and performing electronic searches of internal HBRSEP 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 NOs MC7153 and ME6476) to ensure that any changes being made to ensure 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.

The Fukushima Orders are being independently evaluated. Any plant changes will be evaluated for impact on the fire protection program in accordance with the HBRSEP design change process.

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Duke Energy Attachment P - Rl-PB Alternatives to NFPA 806 10 CFR 50.48(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 HBRSEP.

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Duke Energy Attachment Q - No Significant Hazards Evaluation Q. No Significant Hazards Evaluations 4 Pages Attached Page Q-1 HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page Q-1

Duke Energy Attachment 0 - No Significant Hazards Evaluation A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety Duke Energy 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."

To the extent that these conclusions apply to compliance with the requirements in NFPA 805, these conclusions are based on the following NRC statements in the Statements of Consideration accompanying the adoption of alternative fire protection requirements based on NFPA 805.

Criterion 1: Does the proposed change involve a significant increase in the Probabilityor consequences of an accident previously evaluated?

Response: No.

Operation of the H. B. Robinson Steam Electric Plant Unit No.2 (HBRSEP) in accordance with the proposed amendment does not result in a significant increase in the probability or consequences of accidents previously evaluated. The proposed amendment does not affect accident initiators or precursors as described in the HBRSEP Updated Final Safety Analysis Report (UFSAR), nor does it adversely alter design assumptions, conditions, or configurations of the facility, and it does not adversely impact the ability of structures, systems, or components (SSCs) to perform their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the way in which safety-related systems perform their functions as required by the accident analysis. The SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions.

The purpose of this amendment is to permit HBRSEP to adopt a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance contained in Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection HBRSEP LAR Rev 0 Page Q-2

Duke Energy Attachment Q - No Significant Hazards Evaluation requirements that are an acceptable alternative to the 10 CFR Part 50, Appendix R, fire protection features (69 FR 33536; June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic risk assessments, and fire modeling calculations, have been performed to demonstrate that the performance-based requirements of NFPA 805 have been met.

NFPA 805, taken as a whole, provides an acceptable alternative for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 10 CFR Part 50, meets the underlying intent of the NRC's existing fire protection regulations and guidance, and achieves defense-in-depth along with the goals, performance objectives, and performance criteria specified in NFPA 805, Chapter 1. In addition, if there are any increases in core damage frequency (CDF) or risk as a result of the transition to NFPA 805, the increase will be small, governed by the delta risk requirements of NFPA 805, and consistent with the intent of the Commission's Safety Goal Policy.

Based on the above, the implementation of this amendment to transition the Fire Protection Plan at HBRSEP to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a significant increase in the probability of any accident previously evaluated.

In addition, all 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.

Criterion 2: Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Operation of HBRSEP 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 consequences was included in the evaluation of design basis accidents (DBA) documented in the UFSAR as a part of the transition to NFPA 805. The proposed amendment does not impact these accident analyses. The proposed change does not alter the requirements or functions for systems required during accident conditions, nor does it alter the required mitigation capability of the fire protection program, or its functioning during accident conditions as assumed in the licensing basis analyses and/or DBA radiological consequences evaluations.

The proposed amendment does not adversely affect accident initiators nor alter design assumptions, or conditions of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to maintain the unit in a safe and stable condition remain capable of performing their design functions.

HBRSEP LAR Rev 0 Page 0-3

Duke Energy Affachment 0 - No Sinnificant Hazards Evaluation The purpose of the proposed amendment is to permit HBRSEP 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. As indicated in the Statements of Consideration, 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.

The requirements in NFPA 805 address only fire protection and the impacts of fire effects on the plant have been evaluated. The proposed fire protection program changes do not involve new failure mechanisms or malfunctions that could initiate a new or different kind of accident beyond those already analyzed in the UFSAR. Based on this, as well as the discussion above, the implementation of this amendment to transition the Fire Protection Plan at HBRSEP to one based on NFPA 805, in accordance with 10 CFR 50.48(c), does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion3: Does the proposed change involve a siqnificantreduction in a margin of safety Response: No.

Operation of HBRSEP in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The transition to a new risk-informed, performance-based fire protection licensing basis that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) 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 in the UFSAR to mitigate accidents. The proposed change does not adversely impact systems that respond to safely shut down the plant and maintain the plant in a safe shutdown condition. In addition, the proposed amendment will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without implementation of appropriate compensatory measures. The purpose of the proposed amendment is to permit HBRSEP 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 Regulatory Guide 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 required fire protection features (69 FR 33536, June 16, 2004).

The risk evaluations for plant changes, in part as they relate to the potential for reducing a safety margin, were measured quantitatively for acceptability using the delta risk guidance contained in RG 1.205. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling calculations, have been Page Q-4 HBRSEP LAR Rev 0 LAR Rev 0 Page Q-4

Duke Energy Attachment Q - No Significant Hazards Evaluation performed to demonstrate that the performance-based methods of NFPA 805 do not result in a significant reduction in the margin of safety.

As such, the proposed changes are evaluated to ensure that risk and safety margins are kept within acceptable limits. Based on the above, the implementation of this amendment to transition the Fire Protection Plan at HBRSEP to one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not significantly reduce a margin of safety.

Conclusion The Fire Protection Program in accordance with NFPA 805 will continue 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 risk-informed licensing basis changes, as described in RG 1.174, is consistent with the defense-in-depth philosophy, and maintains sufficient safety margins. Based on the above discussion, the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the amendment request to transition the Fire Protection Plan at HBRSEP to one based on NFPA 805, in accordance with 10 CFR 50.48(c), involves no significant hazards consideration.

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Duke Energy Aftachment R -Environmental Considerations R. Environmental Considerations Evaluation I Page Attached Page Ri HBRSEP LAR Rev 0 LAR Rev 0 Page R-1

Duke Enerciy Attachment R -Environmental Considerations Pursuant to 10 CFR 51.22(b), an evaluation of the license amendment request (LAR) has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c). That evaluation shows that the criteria for a categorical exclusion are satisfied for the following reasons. The LAR does not involve:

A significant hazards consideration.

This conclusion is supported by the determination of no significant hazards consideration.

A 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. This radioactive release goal is to provide reasonable assurance that a fire will not result in a radiological release that affects the public, plant personnel, or the environment. The NFPA 805 transition based on fire suppression activities, but not involving fuel damage, has been evaluated and does not create any new source terms. Therefore, this LAR will not change the types or amounts of any effluents that may be released offsite.

A significant increase in the individual or cumulative occupational radiation exposure.

Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for occupational exposures. Therefore, this LAR will not change the types or amounts of occupational exposures based on the results of the analysis summarized in Attachment E to this document based on fire fighting activities.

In summary, 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.

Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

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Duke Energy Attachment T - Clarification of Prior NRC Approvals T. Clarification of Prior NRC Approvals There are no requests for clarification of prior NRC approvals for the HBRSEP submittal.

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Duke Energy Aftachment U - Internal Events PRA Quality Duke Energy Attachment U Internal Events PRA Quality U. Internal Events PRA Quality 24 Pages Attached Page U-I HBRSEP Rev 0 LAR Rev HBRSEP LAR 0 Page U-1

Duke Enerciv Attachment U - Internal Events PRA Quality In accordance with RG 1.200, Rev. 2 and ASME/ANS RA-Sa-2009, a full scope Peer Review of the internal events model of the HBRSEP PRA was conducted in May 2010 by six PRA experts. The review provided findings and suggestions regarding the model and identified 25 supporting requirements within the internal events portion of the model that did not meet Capability Category (CAT) I1.These findings were either resolved by additional analysis and included in the quantitative results, or evaluated with their impact on the applicable risk evaluation. The dispositions for these findings are presented in Table U-I, Internal Events PRA Peer Review - Facts and Observations.

The HBRSEP internal events PRA model is utilized to calculate CDF/LERF for the NFPA 805 application. Any elements of the supporting requirements detailed in ASME/ANS RA-Sa-2009 that could be significantly affected by the application are required to meet Capability Category II requirements.

The Internal Events PRA provides an adequate base model for the development of the Fire PRA and the NFPA 805 Application. While there has been one instance of a change to the internal events model since the 2010 peer review that could be considered a model upgrade, this change is specifically applicable to the loss of CCW initiating event and does not affect the Fire PRA results.

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Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID)

AS-A5 (1) Define the Accident Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, This is consistent with FRP-H.1 where Step 73a (Cat I-Ill is MET] Sequence Model 4.7, 4.8, & 4.10 correlate safety function and checks for SI pumps running and if not, the procedure success criteria to procedures and supporting requires to go to Step 76 to verify pressurizer PORVs documents; The transient event tree (Figure are all closed. If PORVs are not closed, the procedure 4.3) maps the progression of plant response to directs to close associated PORV Block Valves.

an initiating event as is presented in the ESD. Therefore, description of sequence TBU in RNP-F/PSA-0043 section 4.2.3.5 is re-written to Description of sequence TBU, in 0042, remove assumption.

Section 4.2.3.5, "For sequences involving injection failure, it is assumed that the operators do not open the PORVs following the failure of safety injection. The RCS pressure increases until the pressurizer relief valves lift. The valves cycle to relieve pressure but successfully reclose. The potential for valve failure is addressed in the next sequence." Is this consistent with procedures and bounding for consequences?

AS-A5 (2) Define the Accident Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, Transient-induced RCP seal failure results in transient

[Cat I-Ill is MET] Sequence Model 4.7, 4.8, & 4.10 correlate safety function and similar to an S1 LOCA and is usually limiting in that an success criteria to procedures and supporting S1 LOCA requires inventory makeup and secondary documents; The transient event tree (Figure side heat removal. It is assumed that a small 4.3) maps the progression of plant response to percentage of seal LOCA develops that would require an initiating event as is presented inthe ESD. inventory makeup and heat removal using Safety Injection because of high leak rates, but the large breaks would occur after at least a 15 minute delay Describe how the range of seal LOCA break resulting in reduction in transient severity. Section 4.4A flows, which can exceed small LOCA break of RNP-F/PSA-0043 has been revised to explicitly flow were modeled. include this assumption. Additionally, this assumption and uncertainty is captured in RNP-F/PSA-0074, 'RNP Uncertainty Analysis', for risk impact determination.

HBRSEP LAR Rev 0 Page U-3

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID)

AS-A5 (3) Define the Accident Dispositioned RNP-F/PSA-0043, Section 4.8, Consequential An Induced SGTR model, based on the guidance in

[Cat I-Ill is MET] Sequence Model Pressure-induced SGTR not considered. NUREG-1570, "Risk Assessment of Severe "Consequential tube ruptures due to high Accident-Induced Steam Generator Tube Rupture,"

[Associated with primary-to-secondary differential pressures USNRC, March 1998, was developed and documented AS-B3] (e.g., due to secondary line breaks or ATWS in RNP-F/PSA-0047. Sections 4.8 of RNP-F/PSA-0043 events) are not explicitly considered in the and the Level 2 results have been updated to reflect model. Secondary line breaks can result in a the induced SGTR model and results. More detailed primary-to-secondary differential pressure modeling of Induced SGTR was added to the level 2 equivalent to RCS pressure." analysis as described in section 8.2.2 of RNP-F/PSA-0047, Rev. 2.

AS-B3 (1) Identify Dispositioned Section 7.5 of the system notebook addresses A review of the systems and the impact of the initiating

[Cat I-Ill is NOT phenomenological the requirements of SRs AS-B1 and AS-B3. events on those systems were performed to ensure MET] conditions created by Based on a review of these section of several that the impacts are captured in the model as the accident system notebooks, this reviewer did not appropriate. A statement discussing the impact of the

[Associated with progression identify any components or operator action initiators on the system was added to section 7.1 of the AS-B1 and SY- that have been considered to be impacted by system notebook for each system.

B14] any initiating event (or accident sequences associated with the initiating event). Based on RESOLUTION OF CAPABILITY CATEGORY a discussion with the Duke Energy's Lead CLASSIFICATION PRA engineer, it is understood that the Based on the disposition above, SR AS-B3 is potential impact of an IE on the systems considered to be MET at CAT I-Ill.

credited to mitigate the consequences of the IE have been addressed. However, the discussions provided in the pertinent sections of the system notebooks are not adequate to provide a defensible and easily traceable documentation of the justification of 'no impact.' Given the relatively compact physical structure of the plant, housing of redundant components in the same location, and relatively high dependency on the operator action, a better documentation should be provided to support the 'no impact' assertion.

For example, the potential impact of the SLB outside containment IE on the AFW function and the non-emergency buses should be included. Another potentially vulnerable HBRSEP LAR Rev 0 Page U-4

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) system is the Fire Protection system, where the component of such a system is usually are not credited to work in harsh environment. The system notebook and the plant walk down documentation are thought to provide justification that this issue has been looked at.

Better documentation of the supporting analyses is needed to support the assertion of

'no impact'.

AS-Cl (1) Correlation of system Dispositioned RNP-F/PSA-0043 provides a reasonable For each accident sequence events in

[Cat I-Ill is MET] top events and notebook with clear tables and discussions of RNP-F/PSA-0043 success criteria tables, system top Accident Sequence work performed. However, it is difficult to event gate names are added in the Supporting safety function events correlate between system top events and Documentation column.

accident sequence safety function events.

AS-C2 (1) Correlation of Type B Dispositioned RNP-F/PSA-0006 on Initiating Events RNP-F/PSA-0014 is added to Document Indexing

[Cat I-IIl is MET] & C HRA and identifies transients and LOCAs covered in Table and Section 4.0, References, of Accident Sequence Accident Sequence Analysis. Operator actions RNP-F/PSA-0006.

Analysis are discussed in tables and narrative, but HRA Type B & C Notebook RNP-F/PSA-0014 is not on reference list. A tie to the HRA notebook is needed for clarity.

AS-C2 (2) Accident sequence Dispositioned RNP-F/PSA-0006 on Initiating Events The MTC values selected from WCAP-15831-P and

[Cat I-Ill is MET] development identifies transients and LOCAs covered in listed in Table 4.9 of RNP-F/PSA-0043 for the ATWS Accident Sequence Analysis. Operator actions UET analysis are for a Low Reactivity Core M6del with are discussed in tables and narrative, a maximum HZP MTC of +3.5 pcm/0 F (HFP MTC WCAP-15831-P is used as the basis for equivalent of -7.42 pcm/° F). Per ANP-2887 -cycle Moderator Temperature Coefficient (MTC) reload report for cycle 27, the HBRSEP MTC at HZP is Unfavorable Exposure Time (UET). No -0.35 pcm/0 F at BOC and -11.29 pcm/0 F at HFP. The explanation provided as to how this data WCAP analysis used a bounding MTC that was reflects current plant operation, refueling positive. For HBRSEP, the MTC is negative, therefore cycles, or fuel configuration. bounded by the MTC values of the Low Reactivity Core Model used in the WCAP. Additionally, PSA is on the Reload Review panel and will be evaluating current plant data to existing PSA analysis as indicated in Note HBRSEP LAR Rev 0 Page U-5

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID) under Table 4.9 of RNP-F/PSA-0043.

DA-A2 (1 and 2) Definitions for data Dispositioned 1) SSC boundaries are defined in Consideration of the AFW room cooling as an air handling

[Cat I-Ill is NOT analysis RNP-F/PSA-0077, Attachment 2. No boundary unit does not fit the manner in which the components have MET] definition is defined for HVAC air handling units been modeled. Given the system configuration in the plant or coolers. The AFW coolers are modeled as the separation of the components is more appropriate.

separate fans and heat exchangers. These are NUREG-CRJ6928 provides values for the as modeled more appropriately grouped as an air handling configuration and these are utilized.

unit per NUREG-CR-9628. The failure boundary for the Diesel Generator output breaker is not The definition of the EDG component boundary is correct consistent with the system analysis modeling in that it includes the output breaker. The event and the generic failure data. Attachment 2 has representing the failure of the output breakers the breaker within the component boundary, but (PCB5217BNN and PCB5227BNN) to close will be set to the breaker is modeled separately. Control room zero in the model since failures of this component are hand switches are included in the component rolled into failures of the EDG.

boundary per Attachment 2, even though they are not local as described in NUREG-CR/6968, The control circuitry which is included in the NUREG-this appears to be a grey area, though. Logic for CR/6928 component boundary definitions is interpreted to control valve automation is not within the include the control room hand switches as well as local component boundary, though NUREG-CR/6968 switches. Therefore the definition of component boundaries states local instrumentation and controls are as defined in Attachment 2 of the HBRSEP data calculation within the boundary. For pumps, it appears that RNP-FIPSA-0072 are consistent with the boundary separate cooling and lubrication systems are not definitions identified in NUREG-CR/6928.

within scope, though these systems would be included in NUREG-CRP6928 if local. The peer Component boundaries for control valves as defined in review did not include a comprehensive review Attachment 2 of the HBRSEP data calculation RNP-F/PSA-of component boundaries. The issues found, 0072 are consistent with the boundary definitions identified though individually not a finding, show a trend of in NUREG-CR/6928 in that they specify that local switches, inconsistencies between the plant model and contacts, relays, and control circuitry are included in the component boundary definitions. 2) No boundary. Logic for an automatic actuation is not identified definitions of failure modes were found. 3) No in NUREG-CRP6928 definitions and therefore is defined to definitions of success criteria were found. be outside of the control valve boundary in Attachment 2.

Attachment 2 of the HBRSEP data calculation RNP-F/PSA-0072 specifies that, with regard to pumps, if a different system is providing additional cooling or lubrication then this component is considered to be separate and not within the boundaries of the pump being considered. NUREG-CR/6928 identifies that only local Page U-6 HBRSEPLARRevO HBRSEP LAR Rev 0 Page U-6

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID) lubrication or cooling systems are to be included in pump boundaries and Attachment 2 boundary definitions are in agreement with this specification.

With regard to the battery chargers the input and output breakers are modeled separately due to the fact that plant specific data is utilized in the model for battery chargers.

Thus the manner in which the component is modeled is consistent with those definitions put forth in Attachment 2 of the HBRSEP data calculation RNP-F/PSA-0072.

Generic data values are utilized to model the breakers.

A review was made of other component boundaries and it was determined the boundaries detailed in Attachment 2 of the data calculation (RNP-F/PSA-0072) and in the model are appropriate when compared to those set forth in NUREG-CR/6928.

Wording has been added to Section 4.3 of the data calculation (RNP-F/PSA-0072) to provide more detailed information on the identification of events as failures. There are definitions in the HBRSEP scoping documents that are used in the evaluation of functional failures for the Maintenance Rule. The MR database is then used to capture all failure events. These events are then evaluated and screening comments provided to determine if they count as failures towards the PSA.

Success criteria are comprehensively defined for each modeled system in the respective system notebooks (RNP-F/PSA-0018).

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR DA-A2 is considered to be MET at CAT I-Ill.

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Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID)

DA-Ci (1) Parameter estimation Dispositioned Some generic data collected from a The spreadsheet which documents and aggregates the

[Cat I-Ill is MET] recognized source, NUREG-CR/6928. Other generic data (GENDATA RNP 2011.xlsx) identifies generic data collected from a Progress Energy which data source contributes for each component and source called 'Generic Data Aggregation.' failure mode. This includes the numerical values that Unable to find documentation of process and are used for each source. The Section 3.2 of the data methodology used for data collection and calculation (RNP-F/PSA-0072) provides references for analysis for Generic Data Aggregation.' each contributor to the generic data aggregation.

Generic data is not used for unavailability. A cursory review of Generic Data Aggregation showed that the failure rates were within reason based on similar failure rates in NUREG-CR/6928.

DA-C14 (1) Intersystem Dispositioned No inter-system coincident maintenance PRA personnel reviewed EOOS entries for instances in

[Cat I-Ill is MET] Unavailability analysis was found to be performed. which intersystem coincident maintenance occurred.

While examples of such events could be found, none were the result of planned, repetitive maintenance activities. Therefore, using the guidance of the ASME standard (ASME/ANS RA-Sa-2009) it was determined no additional events should be created for these occurrences.

DA-C3 (1) Plant data screening Dispositioned Plant specific data originates from The wording in Section 4.3 of the data calculation

[Cat I-Ill is MET] Maintenance Rule functional failure data. (RNP-F/PSA-0072) has been adjusted to articulate that Justifications are provided for screening and the MR database is used to capture all failure events.

disregarding plant-specific data in These events are then evaluated and screening RNPFailData.xls. Section 4.3 of comments provided to determine if they count as RNP-F/PSA-0072 states that "Afailure would failures towards the PSA. The statement from Section be counted if a component tripped for no 4.3 of the data calculation quoted in this F&O has been apparent cause and was later restarted with removed from the calculation due to an inaccurate no corrective activity. This type of event may representation of the process used in evaluating not be counted as a MR functional failure..." failures.

However, this type of failure would be screened out by the failure identification process that starts with MR functional failures.

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DA-C4 (1) Plant component Dispositioned For maintenance rule functional failures and This is primarily a documentation issue and wording

[Cat I-Ill is NOT failure data unavailability events, the reviewer was unable has been added to Section 4.3 of the data calculation MET] to find system or component specific (RNP-F/PSA-0072) to provide more detailed definitions of these MR performance criteria, information on the identification of events as failures.

ADM-NGGC-0101 states that PSA should be There are definitions in the HBRSEP scoping considered as an input into MR performance documents that are used in the evaluation of functional criteria development. No evidence was found failures for the Maintenance Rule. The MR database is that MR System functional failure and then used to capture all failure events. These events unavailability definitions have been compared are then evaluated and screening comments provided to definitions or scoping in PSA. This SR is not to determine if they count as failures towards the PSA.

met.

RESOLUTION OF CAPABILITY CATEGORY No clear basis was discovered for CLASSIFICATION identificlation ofsts was diues. e f The documentation has been updated and addressed identification of events as failures. accordingly. Therefore, SR DA-C4 is considered MET at CAT I-Ill.

DA-C6 (1) Plant-specific Dispositioned In practice, PI (computer point) data is used Further documentation has been added to Section 4.4

[Cat I-Ill is MET] component demands for estimating demands on standby of the data calculation (RNP-F/PSA-0072) to components, where the data points exist. demonstrate that post maintenance testing is excluded According to Section 4.4 of RNP-F/PSA-0072, from the collection of demand data. The component estimates of operation based on normal data based on OSI-PI utilizes a one hour screening operating practices, surveillance procedures value. Other demands are based on test procedures or (SPs), and System Engineer input were used system engineer input and only for the remaining AOVs and MOVs, Batteries, include those demands specified during the test or Battery Chargers, Air Compressors, Dedicated during normal operations.

Shutdown Diesel Generator, Diesel Driven Fire Pump, Motor Driven Fire Pump, and Deepwell Pumps. There is no evidence that demands from post maintenance testing are excluded. No unusual number of demands as multiple demands inan hour are excluded.

DA-C8 (1) Component Dispositioned The flag events that represent the time that Data was obtained and analyzed to determine the

[Cat I is MET] configurations in components were configured in their standby runtime percentages of all possible alignment standby status do not appear to use actual plant data configurations for select plant systems. These plant as a primary source. systems were selected based on those for which data was available and those which utilized alignment flags HBRSEP LAR Rev 0 Page U-9

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) in previous model revisions. New basic events were put into the model to replace the former alignment flags. The new events represent each possible running configuration and have a value equal to the percentage that the alignment was found to be operating.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of plant data into this analysis, SR DA-C8 is considered to be MET at CAT Il/111 is met.

DA-D4 (1) Bayesian approach Dispositioned No discussion is provided to demonstrate the Section 5.1 of the data calculation (RNP-F/PSA-0072)

[Cat I is MET] posterior distributions are reasonable given provides identification of ways in which the Bayesian the relative weight of evidence provided by the results have been validated using guidance from the prior and the plant-specific data. standard. Both generic and plant specific data was reviewed and spreadsheet (RNPBayes Data R1.xlsx) calculations analyzed for appropriateness in the performance of the Bayesian update. Results of the update were checked for appropriateness using at least one of the methods put forth in the standard.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of the posterior distributions to the Bayesian approach, CAT Il/111 is considered to be MET.

DA-D6 (1) Common Cause Dispositioned No evidence that a review was performed of Performed a review for plant experience with common

[Cat I is MET] Failure probabilities generic common cause failure probabilities to cause failures, and updated RNP-F/PSA-0073 Section be consistent with available plant experience. 2.0 to include this discussion.

There is no discussion of available plant experience with respect to common cause RESOLUTION OF CAPABILITY CATEGORY failures in RNP-F/PSA-0073. CLASSIFICATION Incorporation of plant experience qualifies for meeting CAT II requirements for DA-D6.

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DA-E2 (1) Data analysis process Dispositioned The PRA documentation does not show the Development and discussion of the occurrences

[Cat I-Ill is MET] documentation probabilities or basis for the probabilities for represented by the basic events cited in the F&O may the following basic events (note that the basic be found in the Westinghouse Owner's Group events are assigned probabilities in the fault document "WOG Risk-Informed ATWS Assessment tree): #ACRDMF (control rod fails due to and Licensing Implementation Process" (WCAP-mechanical binding), #CRDMF (insufficient rod 15831-P) which is cited in the HBRSEP Success insertion), #RPS (failure of reactor trip), Criteria calculation (RNP-F/PSA-0075). Additionally, EAMSAC (AMSAC failure), ESFAS (ESFAS information regarding to the performance of a fails), CCVENT (performing CV purge), and containment purge may be found in the Technical GINRDOORSL (personnel hatch door gasket). Specifications for HBRSEP, Sections 3.3.6, 3.6, and No evidence was found that these basic event 3.9.3.

values are not appropriate.

HR-H1 (1) Operator recovery Dispositioned Section E.5 RNP-F/PSA-0014 Rev. 1 As indicated in Section E.5 of RNP-F/PSA-0014, the

[Cat IIis MET] actions discusses TYPE Cr: POST-INITIATOR only non-proceduralized (Type CR) human interactions RECOVERY ACTIONS; states that the only included in the Robinson PSA model are in the Plant recovery human actions included in the flooding analysis (RNP-F/PSA-0009). OPER-48 and Robinson PSA model are in the Plant flooding OPER-51 are not Type CR.

analysis, but Table A-1 indicates OPER-48 recovery of the loss of IAto the SG PORVs via the use of the Steam Dump N2 accumulator and OPER-51 Failure to recover RCP seal injection with alternate filter line.

Documentation of operator recovery actions is not clear and consistent.

HR-I1 (1) HRA documentation Dispositioned The Type A pre-initiator process is unique and Updated Figure 1-1, Pre-initiator Human Failure Events

[Cat I-Ill is MET] therefore requires some detailed review to Flow Chart of RNP-F/PSA-0071 Rev. 2 to provide identify how HFEs were additional detail about how HFEs are identified.

identified for inclusion in the models.

IE-A6 (1) Initiating events from Dispositioned Several IE's are evaluated with a fault tree that Section 1.1.4.10 and 1.1.4.11 of RNP-F/PSA-0006,

[Cat II is MET] multiple failures models the loss of a system. Such fault trees Rev 3 has been updated to include consideration.

typically include common cause failures and various system alignments caused by test and maintenance unavailability. No consideration is documented for potential unique initiating HBRSEP LAR Rev 0 Page U-11I

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) events for common cause failures of both DC busses or both AC busses. The consequences of such initiating events would be more severe than an initiating event where a single bus fails.

IE-C3 (1) Recovery action credit Dispositioned The screening of the PORV opening due to Should pressure transmitter PT-445 fail high, a

[Cat I-Ill is MET] failure of PT-445 needs to be investigated as a demand for the PCV-456 will be generated. In order for potential unique initiator, the valve to open, the P-11 interlock must be defeated (i.e., RCS pressure > 2000 psi). Should this occur, the opening of the valve will result in a decrease in RCS pressure and a reactor trip on low pressurizer pressure. This event is similar to a small LOCA with a frequency of 5E-4 per year. However, it can be mitigated if the operator closes either the PORV or its associated block valve. When mitigated, it would be similar to the reactor trip initiator of 2.53E-01 per year.

When the potential for operator recovery is included, the initiating frequency for this event is much less than the small LOCA initiating event frequency: From RNP-F/PSA-0072, RNP Failure Data Analysis, the probability of PORV Spurious Operation is 3.31E-07 per hour, PORV Fails to Close is 6.70E-4 per demand, PORV Block Valve Fails to Close is 1.64E-3 per demand and from RNP-F/PSA-0014 the potential for operator recovery is 2.6E-3. The frequency of a spurious PORV opening and failure to reclose is thus 1.5E-5 per year (Note: 2 PORVs) which is small compared to the small LOCA frequency of 5E-4 per year. In addition, it does not result in any other significant impacts on safety systems. Therefore, a spurious PORV opening is not considered to require separate consideration in the HBRSEP PRA model.

RNP-F/PSA-0006, Section 1.1.4.6 is updated to include this justification.

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IE-D3 (1) Initiating event Dispositioned A qualitative assessment of identified IE Rationale for using lognormal distribution is

[Cat I-Il1is NOT analysis uncertainties is presented in documented in RNP-F/PSA-0049, Section 6.5 and MET] RNP-F/PSA-0074. No basis provided for Table 12 note 1 in RNP-F/PSA-0006 is updated selection of uncertainty distributions, accordingly.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated to provide the appropriate basis for this methodology and addressed accordingly. Therefore, SR IE-D3 is considered to be MET at CAT I-Ill.

LE-C 11 (1) Containment failure Dispositioned RNP-F/PSA-0047 credits containment spray Level 2 operator actions were assessed in more detail

[Cat I is MET] impacts on equipment operation following containment failure for using industry-standard methodologies via the EPRI and operator actions small and late releases, but not for large early HRA calculator. As part of that assessment,

[Associated with releases. Manual isolation operator actions adjustments were made as appropriate to the LE-C12] are described in Sections 8.3.6 and 8.3.7. No environmental conditions under the actions that would evidence of justification of credit for equipment be performed. In this context, assessment of survivability or justification of credit for human Containment Spray success is used to determine actions that could potentially be impacted by whether a containment release (non-ISLOCA or SGTR) containment failure, is scrubbed or unscrubbed. In many end-states, in the Containment Event Tree, this determines whether the release is a Large Early release or not. The success or failure of the Containment Spray system to perform its credited functions is determined through the development of failure and unavailability data using industry-approved sources and/or plant-specific data.

Furthermore, the Containment Spray system is an ECCS mitigation system and thus designed to sustain the radiation and temperature effects of post-accident conditions postulated in the design basis. No environmental conditions were identified which required the Containment Spray system to operate beyond their design basis. Thus, it is reasonable to conclude that the Containment Spray system would be available (within the constraints of the failure and unavailability data) during a post-accident containment failure.

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RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With adequate justification for credit of equipment survivability and human actions impacted by containment failure, this SR is considered to be MET at CAT Il/111.

LE-C2 (1) Operator actions Dispositioned LERF-related operator actions following the Level 2 operator actions were assessed in more detail

[Cat I is MET] following onset of onset of core damage are described in the using industry-standard methodologies via the EPRI core damage HRA notebook (RNP-F/PSA-0014, Annex B) HRA calculator.

and RNP-F/PSA-0047. Screening values were set to 0.1 to see if there was sensitivity of RESOLUTION OF CAPABILITY CATEGORY results. For example, operator action to CLASSIFICATION mitigate ISLOCA is evaluated in Section 8.2 of Through analysis of human operator actions with HRA RNP-F/PSA-0047. The probability of failure is calculator as a more thorough approach, CAT Il/111 is assumed to be 0.1 based on "available considered to be MET.

guidance and time available." The scenarios for these operator actions were not adequately documented as consistent with applicable procedures. Need to ensure applicability to the current as built/as operated plant.

LE-C4 (1) LERF model logic for Dispositioned The accident sequences are based on generic Assessment performed in RNP-F/PSA-0062.

[Cat I is MET] accident progression references and on plant specific MAAP analysis. Important mitigation actions in Based on this assessment, and based on the fact that significant accident progression sequences, SI-864A and SI-864B are inseries inthe RWST supply such as SG isolation for SGTR are not line and are in close proximity to one another, it is modeled. Evidence of technical justification for reasonable to conclude that the ISLOCA pathway to demonstrating the feasibility of mitigating the RWST could, at worst case, be isolated by closing actions was not provided. Scrubbing is one valve until it can be closed no further and then discussed for release categories in closing the other until it can be closed no further, and RNP-F/PSA-0048 and brief rationale is so on going back and forth between the valves until the provided when it is credited. Inclusion of line is isolated.

beneficial failures was not observed.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The assessment and justification provided above are HBRSEP LAR Rev 0 Page U-14

Duke Energly Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) considered to meet CAT II for SR LE-C4.

LE-C9 (1) Human actions under Dispositioned Four LERF HFEs documented in Sections 8.2, Level 2 operator actions were assessed in more detail

[Cat I1-111is adverse environments 8.3.3, 8.3.6, 8.3.7, and 8.3.11 of using industry-standard methodologies via the EPRI MET] RNP-F/PSA-0047 are modeled. Evidence of HRA calculator. As part of that assessment, justification for human actions under adverse adjustments were made as appropriate to the environments was not found in the LERF HFE environmental conditions under the actions would be documentation. One of the HFEs, OPER- performed.

ISOL, represents local closure of manual valves given failure of containment isolation.

The environmental conditions associated with these valves could potentially have an impact on the operator's ability to locally close the valves and needs to be addressed. Although this action was not credited, the documentation should be clear about this.

LE-D5 (1) Secondary side Dispositioned Based on a review of a number of calculation Assumption of 10 cycles for SG PORVs is not used in

[Cat I is NOT isolation capability files, SGTR event tree, and fault tree for #RW, the Level 2 analysis. For induced SGTR analysis that MET] analysis it appears that a realistic secondary side was added to the Level 2 analysis, an assumption of isolation capability analysis is performed. 40 cycles was used. Since one stuck-open SG SRV is However, there are a number of important considered sufficient for a large early release, the issues that their resolutions may impact this modeling of multiple SRVs sticking open is extraneous.

conclusion. These issues include a) A justification for the assumed number of cycles RESOLUTION OF CAPABILITY CATEGORY for the PORVs on the faulted SG (which is 10) CLASSIFICATION is not provided. Also, it seems that only 'fail to Based on the above disposition LE-D5 is considered to close' of one SRV is considered in the fault be MET at CAT I1.

tree. b) Additionally, it appears that Operator action to isolate the faulted SG is not included as a potential contributor to the secondary side isolation failure probability. Given that in the current model 1) fail to isolate probability is dominated by the probability of a PORV sticking open, and 2) the SGTR IE is one of the highest contributors to the LERF figure of merit, it is important to ensure that the above apparent issues are fully resolved.

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LE-D6 (1) Thermally-Induced Dispositioned Based on a review of Sections 8.2.4 and 8.3.7 More detailed modeling of Induced SGTR was added

[Cat I is MET] Steam Generator of RNP-F/PSA-0047 a simple representation to the level 2 analysis as described in section 8.2.2 of (SG) Tube Rupture of thermally induced SGTR is performed on a RNP-F/PSA-0047.

Analysis generic basis. Note that 1.200 Gap Analysis comment with respect to ISGTR is a valid RESOLUTION OF CAPABILITY CATEGORY comment. The response to the comment CLASSIFICATION seems to be adequate but could not be Based on the above disposition LE-D6 is considered to verified. be MET at CAT II.

LE-E1 (1) Parameter values for Dispositioned The system level data are documented in the Description of development of input parameters is

[Cat I-Ill is MET] equipment and Level 1 analysis and some Level 2-specific provided in RNP-F/PSA-0047. In many cases, they are operator response parameters are documented in Attachment 2 based on expert judgment.

of RNP-F/PSA-0047. The basis for some of the values was subjective (e.g., see Table 8-

10) and others are based primarily on the IPE submittal and supporting MAAP runs for the Level 2 analysis (e.g., Table 8- 35). Although it is understood that the approach for estimating probabilities for some Level 2 parameters is soft, using IPE results without additional justification may not be as realistic as required for a CC II classification.

LE-F1 (1) Parameter estimates Dispositioned Relative contribution to LERF is presented in Induced SGTR has been added to the Level 2 model

[Cat I1-111 is for accident the quantification notebook for HBRSEP. It is the dominant contributor for large MET] progression (RNP-F/PSA-0077) and the source term bypass containment failure sequences. Dominant phenomena development notebook (RNP-F/PSA-0048). contributors to LERF and containment failure are Induced SGTR contribution was not presented described in sections 3.16 and 3.17 of the among the significant LERF contributor results quantification calculation (RNP-F/PSA-0077).

presented in Section 3.16 of RNP-F/PSA-0077.

LE-F2 (1) Documentation of Dispositioned The updated quantification notebook Review of LERF contributors and results of that review

[Cat I-Ill is NOT Contributors to LERF (RNP-F/PSA-0077) was examined to find are described in section 3.3 of the quantification MET] analysis evidence of a review of the LERF contributors calculation.

for reasonableness. All that was found were entries for "Level 2 HNPSUM Changes" RESOLUTION OF CAPABILITY CATEGORY HEIRSEP LAR Rev 0 Page U-16

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID) toward the end of the Appendix A "Change CLASSIFICATION Log for RNP MOR09." Clear documentation is Based on the above disposition LE-F2 is considered to needed to know what LERF reasonableness be MET at CAT I-Ill.

reviews were performed and what were the results of the reviews to ensure that the SR is fully met.

LE-F3 (1) Documentation of Dispositioned The uncertainty analysis, as documented in Assumptions and Uncertainties related to LERF are

[Cat I-Ill is MET] LERF analysis RNP-F/PSA-0074, contains very little revised and documented in RNP-F/PSA-0074.

documentation of identification of LERF

[Associated with uncertainties. Only about 10% of all the LE-G4] assumptions and uncertainties documented in Table 1 of RNP-F/PSA-0074, seem to be LERF related, even though LERF analysis as whole is supported by many conservative analysis assumptions. Additionally most of the identified uncertainties are not related to the biggest contributor to the LERF figure or merit (i.e., a SGTR event). This reviewer found only one (item 195).

LE-G5 (1) LERF analysis Dispositioned Limitations in the LERF analysis that would RNP-F/PSA-0047, Rev. 2, Section 8.4.2 states that "No

[Cat I-Ill is NOT limitations impact applications are not identified. limitations were identified in Level 2 analysis (including MET] LERF) that would impact applications." However, any assumptions and uncertainties identified in the analysis are addressed in RNP-F/PSA-0074, "RNP Uncertainty Analysis".

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR LE-G5 is considered to be MET at CAT I-Ill.

QU-B3 (1) Truncation analysis Dispositioned The current truncation levels of 1E-10 for CDF An appropriate lower truncation is used in

[Cat I-Ill is NOT and LERF are insufficient to estimate the RNP-F/PSA-0077, Rev. 2, and Section 2.2 truncation MET] mean CDF or LERF required by QU-A3 and study shows convergence for both CDF and LERF.

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RNP-F/PSA-0077 shows that truncation RESOLUTION OF CAPABILITY CATEGORY occurs for CDF, but does not establish that it CLASSIFICATION occurs at 1E-10. The study does not show Based on the above disposition, SR QU-B3 is convergence for LERF. SR QU-B3 describes considered to be MET at CAT I-I11.

an acceptable process for demonstrating convergence, where a successive decade reduction in truncation results in a CDF or LERF change of less than 5%. The HBRSEP results show a decade reduction from 1E-10 to 1E-11 results in a change of approximately 10% for CDF and 30% for LERF, which does not meet the example process in the SR.

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QU-D1 (1) Review significant Dispositioned Documented cutset reviews were inadequate. The following has been added to section 3.3 of the

[Cat I-Ill is NOT cutsets to verify Section 3.3 of RNP-F/PSA-0077 explains that quantification calculation (RNP-F/PSA-0077): Due to MET] model logic the top 50 cutsets were reviewed for CDF and the low truncation settings and large quantity of cutsets LERF. Section 1.1 defines a significant cutsets generated it is not practical to evaluate each cutset in as one of the top 95% of cutsets or a cutset the 95th percentile. (Approximately 3800 cutsets that contributes more than 1% to CDF. The comprise the top 95 %.) All of the top fifty cutsets were top 50 cutsets for CDF are near 50% for all reviewed followed by a review of select cutsets at CDF. LERF cutsets are assumed to have a varying intervals leading up to the 95% level. The top similar limitation. Review of non-significant fifteen cutsets represent 50% of the CDF, and cutset cutsets appeared to be adequate. 17 is the point at which each subsequent cutsets RNP-F/PSA-0077 states that 'A sample of contribute less than 1% to CDF. Cutsets at the bottom non-significant cutsets listed in Table B-2 of the top 50 each only contribute approximately 0.2%

(lower than the 95% of the CDF) was reviewed to total CDF. In addition to the top 50, the bottom 50 for reasonableness and to ensure recoveries cutsets were evaluated to increase confidence that the were being applied.' generated MOR12 was yielding logical results. A list of the 50 top core damage cutsets is provided in Table B-1 in Appendix B. These cutsets are presented in order of descending frequency. A sample of non-significant cutsets listed in Table B-2(lower than the 95% of the CDF) was reviewed for reasonableness and to ensure recoveries were being applied. Part of this review for reasonableness was reviewing which accident sequences were contributing to CDF and LERF, comparing those results to the previous model with consideration of what model changes were made, and performing sensitivity testing as needed to determine the reason for the differences between the previous model and this one.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D1 is considered to be MET at CAT I-I11.

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QU-D3 (1) Review cutsets for Dispositioned Section 3.3 of RNP-F/PSA-0077, Cutset Section 3.3 of RNP-F/PSA-0077, states that flag

[Cat I-Ill is NOT flag events, mutually Results Review (Significant & Non-significant) events settings, mutually exclusive event rules and MET] exclusive events, and indicates that "a panel of PRA experts recovery rules are reviewed for correctness and logical recovery rule reviewed the cutsets for modeling accuracy, results.

accuracy operational consistency and to ensure the correct recoveries were being applied." But RESOLUTION OF CAPABILITY CATEGORY only found limited evidence that results were CLASSIFICATION reviewed to determine that flag event settings and mutually exclusive event rules yield logical results. Found model changes listed in Table Based on the disposition above, SR QU-D3 is A-1 of RNP-F/PSA-0077 that indicated that considered to be MET at CAT I-III.

results had been reviewed.

QU-D4 (1) Compare results to Dispositioned RNP-F/PSA-0077 Section 3.4, Similar Plant Section 3.4 of RNP-F/PSA-0077, is updated to provide

[Cat I is MET] similar plants Review, Table 7 shows a comparison of adequate comparison. Comparison with the LERF was Robinson CDF and LERF values to Turkey conducted in the cut-set review process. However, this Point and Beaver Valley. Plant systems detailed review was not documented in the current differences are compared. However, the MOR.

comparison is inadequate and sources of specific differences (e.g., LOCA contribution) RESOLUTION OF CAPABILITY CATEGORY are not identified. CLASSIFICATION Based on the above disposition, SR QU-D4 is considered to be MET at CAT I1-111.

QU-D7 (1) Review Dispositioned Importance was calculated at basic event Section 3.13 of RNP-F/PSA-0077, documents

[Cat I-Ill is NOT component/basic level, but could not be found at component Component Importance.

MET] event importance level. Review did not find documentation to substantiate review of component importance. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D7 is considered to be MET at CAT I-Il1.

QU-F2 (1) Model integration Dispositioned Several items identified as typically 1. System successes are listed in the system notebook

[Cat I-Ill is NOT process documented for SR QU-F2 were not calculation (RNP-F/PSA-0018) and described in the MET] adequately documented. Missing or success criteria notebook (RNP-F/PSA-0075).

inadequate documentation was identified for 2. A discussion of the sensitivity of accident sequences the following items: 1) description of the to operator actions has been added to section 2.7 of HBRSEP LAR Rev 0 Page U-20

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status FindinglObservation Disposition ID) process used to account for system the quantification calculation (RNP-F/PSA-0077).

successes, 2) discussion of human actions 3. Asymmetries for HBRSEP are discussed in the that are the key factors in causing the system notebook calculation (RNP-F/PSA-0018).

accidents to be non-dominant, 3) discussion of Asymmetries exist in particular on the AC Power asymmetries in the model, 4) list of mutually system (and associated systems like DGs and DC exclusive results, 5) cutset review process, power).

and 6) discussion of RRW for equipment or 4. Mutually-exclusive events are listed in the operator errors. RNPMTX_12.xls file, and there is a cutset file which lists all the mutually-exclusive combinations in the fault tree. Both of these files are part of the quantification calculation (RNP-F/PSA-0077).

5. Cutset review process is described in section 3.3

6. No specific standard requirement to discussion review of RRW, but the review of F-V importance is discussed in the quantification calculation and this is considered adequate for RRW as well.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-F2 is considered to be MET at CAT I-Ill.

QU-F3 (1) Document significant Dispositioned Important basic events are documented in Discussion of top accident sequences for SGTR,

[Cat I1-111 is contributors to CDF Table 10 of RNP-F/PSA-0077. Important IE ATWS and LOCA have been added to section 3.15 of MET] results importances are documented Figure 18. the quantification calculation (RNP-F/PSA-0077).

Detailed accident sequence descriptions are provided in Section 3.15. Some, but not all, of these accident sequences describe the top sequences. For example, no top sequence was described for SGTR, ATWS, or LOCA.

Top event tree sequences are discussed in general, but the top cutsets within the event tree sequences are not described. Detailed description of significant accident sequences or functional failure groups could be improved to provide more useful detail for all significant contributors to CDF.

HBRSEP LAR Rev 0 Page U-21

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID)

QU-F5 (1) Document limitations Dispositioned Some software and hardware limitations are Section 1.3 of RNP-F/PSA-0077, describes limitations.

[Cat I-Ill is MET] of the quantification cited in RNP-F/PSA-0077 Section 2.2, but are The limitations of the hardware and software have process not described, been discussed in section 2.2. One example of limitations is documented as being the truncation level and a discussion is included as to how better confidence is obtained inthe truncation level.

QU-F6 (1) Document Dispositioned Significant' definitions are in Section 1.1 of Justification for not performing a review of all cutsets in

[Cat I-Ill is MET] quantitative definitions RNP-F/PSA-0077. However, the definitions the top 95% was provided in section 3.3 of the for significant basic were not consistently applied. The cutset quantification calculation (RNP-F/PSA-0077).

events, cutsets and results review documented in Section 3.3 of accident sequences RNP-F/PSA-0077 was not performed for the top 95%, as discussed in SR QU-D1. No justification provided for not performing a review of significant cutsets. Also, as discussed in QU-D1, it was not clear that RNP-F/PSA-0077 Table B-4 cutsets included those lower than 95% of the LERF.

SC-Al (1) Definition of core Dispositioned The success criteria definitions are Success criteria definitions documented in calculations

[Cat I-IlI is NOT damage documented in RNP-F/PSA-0075, RNP-F/PSA-0075, RNP-F/PSA-0043, MET] RNP-F/PSA-0061, RNP-F/PSA-0043 and RNP-F/PSA-0061 and RNP-F/PSA-0049 have been RNP-F/PSA-0049. There is some revised to provide consistent statements throughout.

[Associated with inconsistency inthe success criteria definitions Definition for "briefly exceeding 1800 'F" has been SC-A2] used for MAAP analyses. revised to read "ifthe temperature exceeds 1800 *F for a short time (i.e., -1 min)" in RNP-F/PSA-0043 and RNP-F/PSA-0049 based on the MAAP case MLOCA_8-HPI.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the disposition above, SR SC-Al is considered to be MET at CAT I-Ill.

HBRSEP LAR Rev 0 Page U-22

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID)

SC-B3 (1) Definition of Success Dispositioned HBRSEP Success Criteria calculation The success criteria selected for the PSA for the

[Cat I-IlI is MET] Criteria RNP-F/PSA-0075 and Accident Sequence LBLOCAs is based on the success criteria specified for calculation RNP-F/PSA-0043 references the design basis LOCAs in the UFSAR. This includes the

[Associated with supporting thermal hydraulic analyses and/or use of a single train of equipment for injection and SC-C2] other applicable analyses. MAAP was used to recirculation, and injection from the 2 accumulators on confirm the success criteria for large break the non-broken loops. To provide a sense of assurance LOCA in the PSA, which is based on the that the UFSAR success criterion was applicable to the success criteria specified for design basis PSA, the MAAP code was looked at but not credited to LBLOCAs in the UFSAR, according to evaluate LBLOCA scenarios at HBRSEP. Last RNP-F/PSA-0075, Section D.3.1. Although the paragraph in Section D.3.1. of RNP-F/PSA-0075 is LBLOCA MAAP run was described as only a re-written to clarify this.

confirmation of the assumed success criteria, MAAP is not appropriate for analysis of large break LOCAs. MAAP does not conserve momentum and can provide misleading unrealistic results for LBLOCA.

SC-C3 (1) Document model Dispositioned Sources of model uncertainty and related The success criteria calc, RNP-F/PSA-0075, has been

[Cat I-Ill is NOT uncertainty assumptions are inadequately identified and reviewed to identify source of model uncertainties and MET] associated with the development of success related assumptions. The identified assumptions and criteria, uncertainties have been added in Table 1 of calc RNP-F/PSA-0074, RNP Uncertainty Analysis.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SC-C3 is considered to be MET at CAT I-Il1.

SY-Al 1 (1) Failures in system Dispositioned Based on a review of a number of system fault A review of the systems revealed that an RWST

[Cat I-Ill is MET] model affecting trees and system notebooks, it appears that rupture event was not included in the model. Event operability most of the pertinent equipment failures are HTKRWSTFN was developed and added to the model.

included in the system models. However, pipe An evaluation of pipe rupture events was performed ruptures are not included as potential using the criteria in SR SY-A15, and it was determined equipment failures based on low probability of that there is adequate justification to discount pipe pipe rupture. This exclusion is acceptable per failure events for those systems where pipe failures SY-A15 but two additional points need to be were previously assumed to have a negligible impact HBRSEP LAR Rev 0 Page U-23

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) covered: 1) pipe failure probabilities for events on risk. A statement to that effect has been added to that result is the total loss of a system need to the applicable system notebooks.

be compared with the random probability of system failure not only one component failure probability, 2) if the same fault tree is used for quantifying the IE frequency, the low failure probability/frequency may not be true any longer.

SY-Al 1 (2) Failures in system Dispositioned RWST tank rupture not accounted for in the Added basic event HTKRWSTFN under gates J150

[Cat I-Ill is MET] model affecting fault tree. and H044 for RWST rupture.

operability SY-A2 (1) Accuracy of system Dispositioned Based on a review of RNP-F/PSA-0018 SR SY-A2 requires the following: "COLLECT pertinent

[Cat 1-Ill is MET] analysis compared to attachments, (e.g., Attachment A.1, A.9, and information to ensure that the systems analysis as-built, as-operated A.5), it is concluded that pertinent data appropriately reflects the as-built and as-operated

[Associated with plant appears to have been used. Additionally, each systems. Examples of such information include system SY-A3] system notebook appears to include a list of P&IDs, one-line diagrams, instrumentation and control all the pertinent references. However, a direct drawings, spatial layout drawings, system operating relationship between the information in the procedures, abnormal operating procedures, system notebook and the source of emergency procedures, success criteria calculations, information is often not provided (e.g., The the final or updated SAR, technical specifications, source document for Figure A.1.1 is not training information, system descriptions and related provided. Or source of information for design documents, actual system operating instrumentation and control is not stated). experience, and interviews with system engineers and operators."

While it could be considered a good practice, there is no standard requirement to provide direct references within the body of the system notebook documentation to support the statements made therein. Additionally, there is no standard requirement to differentiate between 'informational' and direct references. All references are included in section 8.0 of each system notebook appendix.

It is acknowledged that traceability of system engineer HBRSEP LAR Rev 0 Page U-24

Duke Energy Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR (Observation Topic Status Finding/Observation Disposition ID) review comments was deficient. In subsequent system notebook updates, system engineers were assigned NTMs in Passport in which to record their review comments.

As such, no changes are required to meet this SR.

SY-A23 (1) Develop consistent Dispositioned The HBRSEP utilizes CAFTA type codes such Added Section 2.8 Basic Events Naming Convention to

[Cat 1-Ill is MET] model nomenclature that a component failure mode and its RNP-F/PSA-0049.

associated failure rate are applied consistently in multiple systems. Need documentation of the naming convention for basic event names.

SY-A8 (1) Establish component Dispositioned The boundaries of the modeled components The definition of the EDG component boundary is

[Cat I-Ill is MET] boundaries, appear to match the definitions used to correct in that it includes the output breaker. The event definitions and data establish the component failure data. representing the failure of the output breakers However, there is some evidence that the data (PCB5217BNN and PCB5227BNN) to close will be set boundary definitions are not consistent with to zero in the model since failures of this component the definition of component boundary. For are rolled into failures of the EDG. A review was made example, the definition of the component of other component boundaries and it was determined boundary for EDGs does not include output the boundaries detailed in Attachment 2 of the data breaker but the definition of the component calculation (RNP-F/PSA-0072) and in the model are boundary in the data analysis includes output appropriate when compared to those set forth in breaker. NUREG-CRP6928.

SY-B8 (1) Documentation of Dispositioned The spatial hazards that may impact multiple Additional discussion added to Section 3.2 of the

[Cat I-Ill is NOT Spatial Hazards systems or redundant components in the System Notebook Appendices, as appropriate.

MET] same system are not clearly identified.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SY-B8 is considered to be MET at CAT I-Ill.

Page U-25 HBRSEPLARRevO 1HBRSEP LAR Rev 0 Page U-25

Duke Energy Attachment V - Fire PRA PRA Quality Duke Enemy Attachment V Fire

- Quality V. Fire PRA Quality 26 Pages Attached Page V-I HBRSEPLARRevO HBRSEP LAR Rev 0 Page V-1

Duke Energy Affachment V - Fire PRA Quality In accordance with RG 1.205 position 4.3:

"The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard "supporting requirements" important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category Ill, as described in the PRA Standard."

The HBRSEP Combined Internal Events and Fire PRA was peer reviewed during the period of March 2013. The peer review was conducted by a team of industry personnel (utility and vendor). The Westinghouse Owner's Group performed the review and has documented the outcome via LTR-RAM-13-06 "Fire PRA Peer Review of the H. B.

Robinson Nuclear Plant Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard". As noted in LTR-RAM-13-06, the HBRSEP Fire PRA was found to be consistent with the ASME/ANS Standard and is suitable for supporting risk-informed applications.

The peer review team noted a number of Facts and Observations (F&Os). As documented in LTR-RAM-1 3-06, 85% of the Supporting Requirements (SRs) were assessed at Capability Category II or higher. Approximately eighteen Finding level and nine Suggestion level F&Os were identified during the peer review conducted in March 2013. Duke Energy recognized that the Core Damage and Large Early Release Frequencies were relatively high, as noted in LTR-RAM-13-06. Based on the CDF and LERF values at the time of the initial peer review, coupled with the number of findings associated with the Fire Scenario Selection (FSS) Technical Element (18), Duke Energy decided to have a focused peer review.

The focused peer review was conducted during the period of July 2013 and evaluated the FSS Technical Element based on refinements to approved methodologies and updated documentation. The focused peer review was conducted by Frederick Mowrer (C P Fire, LLC) and Bijan Najafi (Hughes Associates) and is documented via Hughes Calculation No. 0004-0042-415-RPT-001, Robinson Nuclear Plant Fire PRA Focused Peer Review, Revision 0. As noted in LTR-RAM-13-06 and Hughes Calculation No.

0004-0042-415-RPT-001, the Fire PRA does apply the methodologies outlined in NUREG/CR-6850 correctly, is consistent with the ASME/ANS Standard and is applicable for supporting risk-informed applications. Although several of the initial F&Os were resolved, seven new findings and three new suggestions were identified during the focused-scope peer review.

Table V-1 documents the Finding level F&Os associated with both the initial and focused peer reviews.

HBRSEP LAR Rev 0 Page V-2

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition CF-A1-01 REVIEW the conditional failure Dispositioned RNP-0153 is Attachment 9 of RNP-F/PRA-0094 RNP-0153 (Attachment 7 of P2217-1021-01-03)

(CAT III) probabilities for fire-induced and describes cable failure analyzed for the has been updated to reflect all HBRSEP circuit failures HBRSEP Fire PRA, RNP-0153. The third bullet information. Additionally, Reference 5.1 has been

[Associated and of RNP-0153, Section 3.0 states, "All cables are updated to "RNP Fire Safe Shutdown Program with CF-Al] ASSIGN the appropriate assumed to be Thermoset. This is consistent Database, Rev 19".

industry-wide generic values with the cable specifications for safety related for risk-significant contributors cable at BNP and there are no substantive based on the specific circuit differences in failure probabilities for the two configuration under cable types for the purposes of this analysis."

consideration Reference 5.1 of RNP-0153 is "BNP Fire Safe Shutdown Program Database, Rev. 26." In response to a Peer Review question, it was advised that BNP should be RNP, that cable should be Thermo-plastic rather than Thermo-set and the HBRSEP FSSPMD should be referenced instead.

The determination of meeting the requirements of CF-Al depends on the information in RNP-0153. As presently written, RNP-0153 leads to questions of whether it is really applicable to HBRSEP.

Revise and update RNP-0153. Perform a confirmation that the balance of information in RNP-0153 is valid for H.B. Robinson.

CF-A2-01 CHARACTERIZE the Dispositioned No characterization of the uncertainty associated The basic events associated with hot short uncertainty associated with the with the applied conditional circuit failure probabilities have been assigned an error factor (NOT MET) applied conditional failure probabilities was documented in EPM Report when the combined cutset is created using the probability assigned per CF- P2217-1021-01-01, Robinson Fire PRA UNCERT code. Although important for Ali Quantification Calculation, as captured in EC determining the statistical uncertainty of the PRA Al. 90905. cutsets, the criteria needed for the LAR are based on the mean values which are not significantly impacted by the uncertainty analysis.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

Based on the above disposition, SR CF-A2 is considered to be MET at CAT 1/11/111.

HEIRSEP LAR Rev 0 Page V-3

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition CS-A1-01 IDENTIFY cables whose fire- Dispositioned Attachment 7 of RNP-F/PSA-0066 (Equipment All cables identified in Attachment 7 of induced failure could adversely Selection) contains a list of equipment which is RNP-F/PSA-0066 have been routed and added (CAT 1/11/111) affect selected equipment to be credited in the Fire PRA, but which does to FSSPMD with the exception of components and/or credited functions in the not have cables identified and routed. A spot that were not installed prior to RNP-F/PSA-0066

[Associated Fire PRA plant response check review of the cable database (RNP being completed.

SR: CS-A10, model. FSSPMD R21_Read Only QA Record.mdb)

CS-A2] reveals that several of the components listed in Attachment 7 are not included in the cable database. For example, Attachment 7 includes "Fire Tag" CHG-C-INDICATING-LIGHT corresponding to PRA BE JILCHGPCTF; however, CHG-C-INDICATINGLIGHT is not in the cable database, hence BE JILCHGPCTF will never be affected by fires. Others identified by spot check include 480V-52/11A, 480V-52/13B, EDGA-AMMETER, etc. Thus, there are some PRA components which are being credited in the PRA which do not have cable routing incorporated into the cable database; hence, they will never fail.

Review Attachment 7 of RNP-F/PSA-0066, identify components which are not in the cable database, and update the cable database to include all credited PRA components.

HBRSEP LAR Rev 0 Page V-4

Duke Energ]y Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition CS-Al 1-01 If assumed cable routing used Dispositioned The scope and extent of assumed cable routing As noted in HBRSEP Change Package in the Fire PRA, IDENTIFY the for non-credited components (i.e., components RNP-0152 (Attachment 19 of (CAT 1/11/111) scope and extent, and assumed to fail for all fires) is documented in P2217-1021-01-03), cable toning was used to PROVIDE a basis for the Table 4 of RNP-F/PSA-0066. Although a basis is confirm cable routes. There were some instances assumed cable routing. not provided for the assumed routing, it is a where cable toning was not possible within a generally accepted practice to omit certain specific compartment (embedded cable, etc.). In

[Associated systems which do not perform a safety function these instances, the cable was assumed to be SR: CS-C3] (but may back up a safety system) and would failed throughout the entire compartment that it require immense amount of work to manually was known to traverse through.

trace the cables. The credited components with assumed routing are contained in the FSSPMD, As noted in HBRSEP change package however, there is no documented basis for the RNP-0205, (Attachment 19 of routing, and it is unclear if the scope and extent P2217-1021-01-03), the assumed cable route is understood. The basis for assumed routing is data determines the cable-to-fire zone correlation not documented. Furthermore, to understand, (which is sufficient for NSCA), but does not and to be able to evaluate the uncertainty determine the cable-to-raceway-to-fire zone associated with assumed routing some form of correlation (which is needed for PRA). Based on documentation should be assembled describing this assumption, any ignition source within a the scope and extent of the assumed routing. given fire zone will impact all cables with For example, are there fire compartments with a assumed cable routes in the ignition source's fire significant amount of assumed routing, are there zone.

high significance fire compartments, systems, trains or components dominated by cable failure of cables which have an assumed routing, etc.

As part of an overall Task 3 documentation package, describe the scope and extent of assumed routing used in the HBRSEP Fire PRA.

Page V-S HBRSEP LAR Rev 00 Page V-6

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition CS-Cl-01 DOCUMENT the cable Dispositioned There is no notebook encompassing Task 3 In the process at HBRSEP, Fire Protection/NSCA selection and location (Cable Selection) making review and update develops and maintains the cable selection and (NOT MET) methodology applied in the Fire difficult. There are numerous change packages, circuit analysis data. This data is then referenced PRA in a manner that and a database (FSSPMD) which is a repository as inputs to the Component Selection and facilitates Fire PRA for the cable routing information; however, there Quantification FPRA calculations. This process applications, upgrades, and is no document explaining what tasks were and associated results are easily reviewable, has peer review. performed, which procedures or guidelines were been peer reviewed multiple times for our other employed, and in which document the analysis is sites and found to be acceptable. There is no contained. requirement to have a separate PRA notebook.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

With no change being made, HBRSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR CS-Cl to be assessed as CAT 1/11/111 is MET.

Page V-6 HBRSEP LAR Rev 00 LAIR Rev Page V-6

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition ES-A2-01 REVIEW power supply, Dispositioned Power supplies, support systems, interlock A detailed review of the modeling of associated interlock circuits, circuits and instrumentation are included in the circuits for all components functional states in the (NOT MET) instrumentation, and support internal events PRA model. For fire PRA specific fire PRA CAFTA model was performed. The system dependencies and logic, examples indicate that not all support details of the review are provided in Attachment 2

[Associated IDENTIFYadditional system adliexc dependencies dependencies t ewerenotall IDENTIFY d additional system completelyt of Hughes 1RCS04042.414.031-002,Associates Inc.

and revisions to(HAl) the SR: ES-B4, equipment whose fire-induced considered. The area of concern is the lack of fire PRA CAFTA model, the RR file table BE, and CS-A2] failure, including spurious modeling of instrumentation power dependency Component Selection Database (CSDB) table actuation, could adversely (self-identified by the utility). For example, see PRASSEL as a result of this detailed review are affect any of the equipment gate HRAPORV-455C "PORV FAILS TO listed in 1RCS04042.414.031-001, Attachments identified per SR ES-Al. CLOSE DUE TO HRA", one of its inputs is an 3, 4, and 5, respectively. Revision 4 of the AND gate for "INDICATIONS THAT PORV IS HBRSEP fire PRA calculation selection (RNP-F/PSA-0066) for component incorporates all of OPEN FAIL." For the indicators modeled here, if the changes required to the fire PRA CAFTA there is a loss of power, the indicators will scale model as a result of the detailed review of the low, potentially preventing the operators from associated circuits.

taking necessary actions. For the indication discussed in the example above, a loss of power Review credited components to ensure that will cause the indicators to scale low, potentially power supplies, support systems, interlock preventing the operators from taking necessary circuits, adequatelyandcaptured.

instrumentation dependencies are impact fire actions, and thus could adversely risk. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

With the above described changes incorporated, this SR is considered to be MET at CAT I/Il/Ill.

Page V-7 HBRSEPLARRevO HIBIRSEP LAR Rev 0 Page V-7

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition ES-A3-01 INCLUDE equipment whose Dispositioned The "RCP Seal No. 1 Leak-Off and "SI test line" RNP-0148, Multiple Spurious Operation Expert fire-induced failure, not are both ISLOCA paths that were screened from Panel Report, did consider spurious isolation of (NOT MET) including spurious operation, the Internal Events analysis; however these the RCP Seal No. 1 leakoff valves. As noted in contributes to or causes screened paths are not explicitly dispositioned in the report, 125V DC power will be removed within (a) fire-induced initiating the fire PRA component selection calculation.

events treated in the Fire Safe These items were discussed with the host utility ten minutes as noted in HBRSEP Dedicated Shutdown / Appendix R that provided the following input on the two Shutdown Procedure (DSP-002).

analysis ISLOCA paths mentioned above: "The seal leak-(b) Internal Events PRA off is normally open and seal leak-off is part of As noted in Attachment 3 of the HBRSEP initiators as identified using the ofi normal open andsalule is par Component Selection Calculation, RCP Seal No.

IE requirements in Part 2 the normal cooling path. Failure in either 1 Leak-Off is not consequential. Leakoff is not (including any gradations direction is non-consequential" and "The test line required to maintain adequate seal cooling in the across capability categories in has 3+ locked closed manual valves." Based on PRA model. Furthermore, seal leakoff is not that standard) as modified per this feedback, it is judged that this is a modeled in the Fire PRA.

4-2.5, or documentation issue with no impact on the (c) unique fire-induced initiating analysis. This finding is related to the events not addressed or consideration of previously screened ISLOCA RESOLUTION OF CAPABILITY CATEGORY otherwise screened from the pathways from the Internal Events PRA for CLASSIFICATION:

above two analyses if SR inclusion in the Fire PRA. Include, in the IE-C4 in Part 2 cannot be met. component selection calculation, a discussion of Based on the above disposition, this SR is screened paths and why or why not they contain considered to be MET at CAT 1/11/111.

components that warrant inclusion in the fire PRA.

FQ-B1-01 PERFORM the quantification in Dispositioned In review of the fire PRA documentation there no RNP-F/PSA-0077 documented proof of (CAT 1/11/111) accordance with HLR-QU-B convergence study was performed in support of convergence for CDF and LERF for the internal and its SRs in Part 2 the selection of truncation level for quantification. events PRA. As the fire PRA model is largely and dependent on the internal events PRA, it can be DEVELOP a defined basis to concluded that there is convergence for CDF and support the claim of non- LERF in the fire PRA applicability of any of the requirements under HLR-QU-B The model was quantified based on a CDF in Part 2. quantification of 1E-12. This is about seven orders of magnitude below the final CDF. This is generally accepted as a good bounding truncation level HEIRSEP LAR Rev 0 Page V-8

Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition FQ-.F1-01 DOCUMENT the CDF and Dispositioned The contents of the elements of applicable SRs The HBRSEP Fire PRA was developed using the (CAT I/Il/Il) LERF analyses in accordance of Part 2 were addressed in the FQ and Internal Events PRA. The Internal Events PRA is with HLR-QU-F and HLR-LE-G associated documents; however, no explicit aligned with RG 1.200 and was peer reviewed.

and their SRs in Part 2 with the connections were established in the documents Therefore, the "back-references" associated with following clarifications: to associate with the "back-references" requirements LE-G2, LE-G4 and LE-G5 are (a) SRs QU-F2 and QU-F3 of requirements LE-G2, LE-G4 and LE-G5. considered to be met.

Part 2 are to be met including identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors (as defined in Part 1);

(b) SR QU-F4 of Part 2 is to be met consistent with 4-2.13 (c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Part 2 are to be met consistent with 4-2.13, and DEVELOP a defined basis to support the claim of non-applicability of any of the requirements under these sections in Part 2.

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SR Topic Status Finding Disposition FSS-Bl-01 DEFINE and JUSTIFY the Dispositioned The conditions and timing that lead to Main Section 5.9.2 of P2217-1021-01-03 provides a (CAT 1/11/111) conditions that are assumed to Control Room (MCR) abandonment due to discussion of Main Control Room abandonment cause MCR abandonment environmental conditions and reliance on ex- due to environment conditions.

[Associated and/or reliance on ex-control control room actions are described in main SR: FSS-A6, room operator actions including control room analysis report (Report The probability associated with main control room FSS-B2] remote and/or alternate 0004-0042-412-002). abandonment due to a loss of habitability has shutdown actions. been incorporated into the quantification of the Abandonment due to equipment damage and to Fire PRA.

loss of habitability is based on the guidelines provided in Section 11.5.2.11 of NUREG/CR-6850. The analysis method described in NUREG/CR-6850 Appendix L is used to assess scenarios in the MCB.

This SR is considered met. A new Finding (FSS-B1-01) has been assigned to this SR because all conditions requiring reliance on remote/alternate shutdown are not identified.

Fires in the MCR or other location in the plant that may lead to loss of control room functions such that use of remote/alternate shutdown capability is required are not characterized and evaluated. Such scenarios for MCR habitability are identified and analyzed.

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SR Topic Status Finding Disposition FSS-C2-01 CHARACTERIZE ignition Dispositioned Section 5.6.2 of the Quantification Calculation Section 5.6.2 of the Fire PRA Quantification source intensity using a (EPM Report No. P2217-1021-01-01, Rev. 1, Calculation discusses the growth profile used.

(CAT Il/111) realistic time-dependent fire dated February 2013 included in EC 90905) and This can also be seen in the RNP_EVAL_Rev2 growth profile (i.e., a time- in particular the third bullet under Figure 7 on spreadsheet used during the quantification dependent heat release rate) Page 51 of 113 indicates that "The HGL process. This spreadsheet includes a cable tray for significant contributors as threshold assumes a constant HRR up to HGL propagation model (HGLTime worksheet).

appropriate to the ignition formation." The basis for using the time versus source. HGL based on a constant HRR and then Furthermore, Calculation NED-M/MECH-1009 compare it to a curve based on a variable HRR also provides a time to damage based on fire has not been provided. At the top of Page 52 of growth.

113 in the Quantification Calculation it states that, "The net effect of these uncertainties is generally a conservative time to HGL. For ignition sources that are high risk, more detailed fire modeling may be pursued on a case by case basis."

The intent of the Standard is that conservative =

Category I, and that realistic = Category Il/111, therefore this SR is evaluated at Category I. An example of a more realistic analysis is Hughes Report Number: 0004-0042-000-001 for Fire Compartment 20.

Finding FSS-C2-01 has not been resolved.

For each scenario, the ignition source intensity is characterized using a time-dependent heat release rate, consistent with the Category Il/111 requirement for this SR; however, the total heat release required to cause hot gas layer formation is based on a fire that is initiated at full peak intensity, consistent with the Category I requirement for this SR. Suggestion FSS-C2-02 has been prepared to suggest further justification and validation of the methodology used to determine the total heat release required to cause hot gas layer formation.

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SR Topic Status Finding Disposition FSS-C4-01 If a severity factor is credited in Dispositioned The severity factors, as described in Section Section 5.5.4.5 of the Quantification Calculation the analysis, ENSURE that 5.5.4.5 of the Quantification Calculation (EPM (EPM Report No. P2217-1021-01-03) has been (CAT II) Report No. P2217-1021-01-01, dated updated to provide more detail as to why specific

a. the severity factor remains February 2013 included in EC 90905) are severity factors are applied in the MakeSScen independent of other apparently evaluated using the calculation logic module.

quantification factors in Attachment 20 of the Quantification Calculation. This is not explained in a clear and

b. the severity factor reflects concise way in Section 5.5.4.5, nor is any the fire event set used to reference to Attachment 20 found in that section.

estimate fire frequency The general reference to module "MakeSScen" and the fact that its contents are listed in

c. the severity factor reflects Attachment 20 in the introductory paragraph of the conditions and Section 5.5.4 is not considered to be sufficient.

assumptions of the Further, the descriptions in Section 5.5.4.5 specific fire scenarios (pages 41-42 of 113) of how the Severity Factor under analysis, and determination is done do not appear to exactly

d. a technical match with the programming logic found on basis supporting the severity pages 7-12 of Attachment 20. This needs to be factor's determination is clarified and clearly documented to support provided. future use, update, and peer review of the Severity Factor calculations.

In addition, the use of generic fire modeling data and severity factors for different ignition source

'Bins" from NUREG/CR-6850 without considering mode of exposure and position (i.e.,

not just distance) of the targets relative to the fire source may not fully constitute "explicit consideration" in quantifying the severity factor such that it reflects the conditions and assumptions of the specific fire scenario under analysis.

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SR Topic Status Finding Disposition FSS-C7-01 If multiple suppression paths Dispositioned Credit is given for both automatic sprinkler No credit has been applied for manual actuation are credited, EVALUATE and suppression and manual firefighting, but mutual or recovery of fixed suppression systems. The (NOT MET) PROPERLY MODEL dependency on the common water supply common water supply at HBRSEP is sufficient to dependencies among the system has not been evaluated or properly provide water to both the automatic suppression

[Associated credited paths including modeled. system as well as the manual firefighting. The SR: FSS-D6] dependencies associated with reliability/unavailability of the fire pump and recovery of a failed fire associated sprinkler has already been accounted suppression system, if such for in the non-suppression probability.

recovery is credited.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

Based on the above disposition, SR FSS-C7 is considered to be MET at CAT 1/11/111.

FSS-D2-01 USE fire models that have Dispositioned This finding recommends that the HGL Attachment J of the LAR provides a discussion sufficient capability to model calculation (RNP-M/MECH-1826) be subjected on the software used during the development of (CAT I/Il/Ill) the conditions of interest and to validation and verification in order to establish the HBRSEP HGL Calculation only within known limits of its technical basis and known limits of (RNP-M/MECH-1826).

(Associated applicability. applicability.

SR: FSS-D6]

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SR Topic Status Finding Disposition FSS-D7-01 In crediting fire detection and Dispositioned There is a failure to meet the Category I Currently, system performance is monitored and (NOT MET) suppression systems, USE requirement of having systems installed and maintained at a high level as part of the System generic estimates of total maintained in accordance with applicable codes Health Reporting and System Notebook system unavailability provided and standards. The Main Turbine Lube Oil processes. Outlier behavior with respect to that Deluge system must be replaced to account for system availability would be evident to the (a) the credited system is system deficiencies identified in NCR-425437 system engineer and plant management through installed and maintained in where a simultaneous actuation of the Turbine the health data (available for the previous accordance with applicable Lube Oil suppression system, along with the 12 months), which indicates overall Excellent codes and standards mezzanine and ground level sprinkler systems, (Green) performance.

(b) the credited system is in a could place a higher system demand on the fully operable state during plant water supply than can be provided by a single Furthermore, during the periods when key fire operation, and fire pump. protection systems are unavailable due to testing (c) the system has not and maintenance, compensatory actions are experienced outlier behavior This was not identified, nor is a comparison taken such that the risk associated with the relative to system provided in the Fire PRA of all installed detection system being unavailable does not increase.

unavailability. and suppression systems vs. the corresponding Code Compliance calculation. RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

Finding FSS-D7-01 has not been resolved.

Because evidence is not provided to support that Using plant-specific information to quantify total credited detection/suppressions systems are unavailability factors is a CAT III requirement and installed and maintained in accordance with was not done.

applicable codes and standards. System health report for period Q2-2013 for systems With no change being made, HBRSEP considers 6185/6181/6175/6195/6205/6180 notes that age, the risk results from the Fire PRA to be creditable obsolescence and replacement part for the NFPA 805 application and this finding to procurement is an issue. This system health be sufficiently resolved for SR FSS-D7 to be report also notes that "There are LTAMs assessed as CAT II is MET.

budgeted for 2014 and 2015 which study and replace the detection, C02, and Halon Systems." This report suggests that some of the fire protection systems at HBRSEP may be experiencing outlier behavior relative to system unavailability and may not be in a fully operable state during plant operation. Consequently, this SR is still considered to be not met.

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Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition FSS-D7-03 In crediting fire detection and Dispositioned Evidence needs to be provided to support that Currently, system performance is monitored and (NOT MET) suppression systems, USE credited detection/suppressions systems are maintained at a high level as part of the System generic estimates of total installed and maintained in accordance with Health Reporting and System Notebook system unavailability provided applicable codes and standards. System health processes. Outlier behavior with respect to that (a) the credited system is system availability would be evident to the installed and maintained in report for period Q2-2013 for systems system engineer and plant management through accordance with applicable 6185/6181/6175/6195/6205/6180 notes that age, the health data (available for the previous 12 codes and standards, and (b) obsolescence and replacement part months), which indicates overall Excellent the credited system is in a fully procurement is an issue for multiple fire (Green) performance.

operable state during plant protection systems. This system health report operation. also notes that "There are LTAMs budgeted for Furthermore, during the periods when key fire are4 buydgaethed f protection systems are unavailable due to testing 2014 and 2015 which study and replace the and maintenance, compensatory actions are detection, C0 2, and Halon Systems." This report taken such that the risk associated with the suggests that some of the fire protection system being unavailable does not increase.

systems at HBRSEP may be experiencing outlier behavior relative to system unavailability and RESOLUTION OF CAPABILITY CATEGORY may not be in a fully operable state during plant CLASSIFICATION:

operation. Using plant-specific information to quantify total unavailability factors is a CAT III requirement and was not done.

With no change being made, HBRSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-D7 to be assessed as CAT II is MET.

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SR Topic Status Finding Disposition FSS-El-01 For any fire modeling Open Section 4.3 of Calculation No. P2217-2100-00, Section 4.3 of Calculation P2217-2100-01-01 will (CAT 1/11/11) parameters not covered by the Fire Scenario Data, RNP-F/PSA-0079, dated be updated appropriately at a later date. This is a requirements of HLR-FSS-C or January 2013 contains information about fire documentation issue that will not impact fire HLR-FSS-D, USE plant- modeling parameters that were used. However, scenario development or quantification.

specific parameter estimates forcifire paamodeln if iable, for fire modeling if available, oSection or 4.4 through 4.7 should be completed use generic information because they are missing information about modified as discussed in SR other relevant fire modeling parameters.

FSS-E2; USE generic information for the remaining Add relevant information to the report.

parameter estimates.

Finding FSS-EI-01 has not been resolved.

Sections 4.3 through 4.7 still make reference to databases for the parameters used in the fire modeling. These parameters should be added to the report.

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SR Topic Status Finding Disposition FSS-E3-01 PROVIDE a mean value of, Open No statistical representation of uncertainty HBRSEP used the HRRs and applied them using (CAT I) and statistical representation intervals (e.g., NUREG/CR-6850 Table E-1 or the guidance found in NUREG/CR-6850. As of, the uncertainty intervals for G-1 for HRR, Tables E-2 through E-9 for severity NUREG/CR-6850 is the consensus methodology, the parameters used for factor) is documented for the mean values of a detailed uncertainty analysis on these modeling scenarios, the significant fire parameter estimates used for fire modeling the parameters is not needed and does not add to significant fire scenarios, the credibility of the results. The majority of applied values are based on the 98th and 75th percentile fires from NUREG/CR-6850, and the ZOIs are applied conservatively. It is not believed that reducing these values would allow the use of reduced impacts for the applications being pursued.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

Although no change has yet been made that would improve the Capability Category assessments, HBRSEP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics.

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SR Topic Status Finding Disposition FSS-E4-01 PROVIDE a characterization of Dispositioned There is some assumed cable routing applied As noted in HBRSEP Change Package (NOT MET) the uncertainties associated directly in FSSPMD. These routings are RNP-0152 (Attachment 19 of with cases where cable routing identified in RNP-0152. Components with P2217-1021-01-03), cable toning was used to

[Associated has been assumed based on unknown routing are otherwise assumed to fail in confirm cable routes. There were some instances SR: FSS-A3] SRs CS-A10 and/or CS-Albl. thenown rodel. ar, there is no diluin where cable toning was not possible within a the PRA model. However, there is no discussion specific compartment (embedded cable, etc.). In that characterizes the uncertainties associated these instances, the cable was assumed to be with cases where cable routing has been failed throughout the entire compartment that it assumed. A more detailed characterization of was known to traverse through.

the uncertainties associated with cases where cable routingto has been assumed is needed for As noted in HBRSEP change package RNP-0205 SR FSS-E4 be met. (Attachment 19 of P2217-1021-01-03), the assumed cable route data determines the cable-to-fire zone correlation (which is sufficient Provide a detailed uncertainty characterization for NSCA), but does not determine the discussion in appropriate reports related to cable-toraceway-to-fire zone correlation (which circuits with assumed routing. is needed for PRA). Based on this assumption, any ignition source within a given fire zone will impact all cables with assumed cable routes in the ignition source's fire zone.

RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:

Based on the above disposition, SR FSS-EF-04 is considered to be MET at CAT I/Il/Ill.

FSS-F3-02 In crediting fire detection and Dispositioned Section 2.4 of the Structural Steel report The unavailability of the Turbine Lube Oil Deluge (CAT 1/11/111) suppression systems, USE (P2217-2300-01-03) states that "It is assumed system has been updated to 0.05 in Rev. 3 of generic estimates of total that an unavailability value of 0.01 will be P2217-1021-01-03 (Fire PRA Quantification

[Associated system unavailability provided bounding and conservative for the deluge Calculation). This change was made based on SR: FSS-D7] that (a) the credited system is installed and maintained in sprinkler system." What is the basis for this? Has engineering judgment.

accordance with applicable this value been confirmed against the plant-codes and standards, and (b) specific experience for availability of the the credited system is in a fully detection/suppression systems?

operable state during plant operation.

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SR Topic Status Finding Disposition FSS-G1-01 APPLY all the supporting Dispositioned Fire modeling performed in support of the Multi- A detailed multi-compartment analysis was (NOT MET) requirements listed in SRs Compartment Analysis (MCA) failed to consider performed and is documented in FSS-Cl through FSS-C8 for the acute effect of hot gas flow through openings RNP-F/PSA-0089. The Fire PRA Quantification

[Associated fire modeling of single physical and ducts on local potential targets. In addition, Calculation only updated results based on new SR: FSS-G2, FSS-G3] analysis units to the modeling the consideration and evaluation of additional CCDPs and a new HGL Calculation. The open of multi-compartment fire aspects of multi-compartment fire scenarios was Turbine Building was discussed in the HGL scenarios, not documented and/or done, i.e., the fact that Calculation (RNP-M/MECH-1826). The detailed the Turbine Building has no exterior walls was review of hot gas flow through openings has not not identified and the impact of this evaluated, been performed. The impact of hot gas flow Treatment of MCA in only 7 pages of the through openings and ducts on local targets is Quantification Calculation does not appear to be expected to be minimal.

adequate. Furthermore, in their evaluation of compartments subject to a hot gas layer, RESOLUTION OF CAPABILITY CATEGORY HBRSEP used a criterion of 1E-07 for total CLASSIFICATION:

ignition frequency. Based on this, HBRSEP Based on the above disposition, SR FSS-GI-O1 excluded all but two compartments. A review of is considered to be MET at CAT the table that was subject to the review showed there were two additional compartments that met the criterion for inclusion. This is an error rate of 100%.

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Duke Energy Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition FSS-G6-02 QUANTIFY the risk Dispositioned Table B-5 of this report defines lists fourteen (14) Section 5.10 of P2217-1021-01-03 provides a (CAT I) contribution of any selected unscreened multi-compartment fire scenarios, discussion of the multi-compartment scenarios multi-compartment fire Attachment E of the same report offers risk quantified in the Fire PRA. There are five scenarios in a manner associated with these scenario. The scenarios that do not screen currently based on consistent with the EQ Quantification Report (P2217-1021-01-02) an updated HGL Calculation Section 5.10 states that "Based on updated Hot (RNP-M/MECH-1826, Rev. 0). These scenarios Gas Layer frequencies, all of the fire have been quantified in a manner consistent with compartments previously analyzed in the Multi- the FQ requirements.

Compartment Analysis (Reference 3.14), have now screen." Reference 3.14 is RESOLUTION OF CAPABILITY CATEGORY RNP-F/PSA-0089 Rev 0 and it does not support CLASSIFICATION:

the statement that all MC fire is screened. Based on the above disposition, SR FSS-G6 is Finding FSS-G6-01 has been resolved. Hughes considered to be Met at CAT Il/Ill.

Calculation No. 1 RCS04042.414.031, Revision 0 (Multi-Compartment Analysis for the Fire Probabilistic Risk Assessment at Robinson Nuclear Plant), provides a detailed multi-compartment analysis. This analysis has been included in Section 5.10 of the HBRSEP Fire PRA Quantification Calculation. Based on updated Hot Gas Layer data, the scenarios identified in the Hughes calculation screen from further analysis.

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SR Topic Status Finding Disposition FSS-H2-01 DOCUMENT a basis for target Dispositioned The fire modeling in Hughes Report No. Plant modifications in addition to intumastic cable (CAT I) damage mechanisms and 0004-0042-000-001 for Fire Compartment 20 coating are incorporated into the Fire PRA thresholds used in the analysis, credits fire coating. Category II of this SR Calculation (EPM Report No.

including references for any requires documentation of the references for any P2217-1021-01-03). These plant modifications plant-specific or target-specific are discussed in Attachment S of the LAR. The performance criteria applied in plant-specific or target-specific performance ten minute time until cable damage (per the analysis. criteria applied in the analysis, and a basis for NUREG/CR-6850) is achieved via plant target damage mechanisms and thresholds used modifications discussed in Attachment S.

in the analysis, which has not been provided.

Finding FSS-H2-01 has not been resolved. RESOLUTION OF CAPABILITY CATEGORY Plant-specific documentation should be provided CLASSIFICATION:

for the performance criteria used to evaluate nd abl trys ithBased on the above disposition, SIR FSS-H2 is damae t cotedcabes damage to coated cables and cable trays with considered to be Met at CAT Il/111.

solid bottoms.

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Duke Energy Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition HRA-C1-01 For each selected fire scenario, Dispositioned The fire-specific operator actions are evaluated All fire response actions have been updated such QUANTIFY the HEPs for all using only the Cause-Based Decision Tree that they are quantified using the CBDTM and (CAT II) HFEs and ACOUNT FOR method (CBDTM). The HRAs from the Internal HCR/ORE combination method similar to Internal relevant fire-related effects Events methods used both CBDTM and Events operator actions. P2217-1022-01-03

[Associated using detailed analyses for HCR/ORE to address the cognitive risk of the documents the Fire HRA.

with FQ-C1] significant HFEs and operator actions. The HCR/ORE method is conservative estimates (e.g., generally the dominant risk value for actions with screening values) for non- short system time windows or very long median significant HFEs, in response times. Events OFIREOMA01 and accordance with the SRs for OFIREOMA02 are events with short time HLR-HR-G in Part 2 set forth windows where the HCR/ORE method would be under at least Capability the dominant cognitive risk. The cognitive risk is Category II, with the following underestimated by an order of magnitude.

clarification:

The HRA needs to be revised to better address

a. Attention is to be given to the cognitive risk portion for each HEP.

how the fire situation alters any previous assessments in non-fire analyses as to the influencing factors and the timing considerations covered in SRs HR-G3, HR-G4, and HR-G5 in Part 2 and

b. Develop a defined basis to support the claim of non-applicability of any of the requirements under HLR-HR-G in Part 2.

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SR Topic Status Finding Disposition IGN-A7-01 USE a plant-wide consistent Dispositioned In general the SR was met. There are areas Transient fire scenarios have been postulated in methodology based on such as the transient ignition sources related each fire compartment as noted in Hughes (CAT 1/11/111) parameters that are expected identification are still required, some of these is Calculation No. 1RCS04042.414.031-002.

to influence the likelihood of already self-identified as part of the "lead review" ignition to apportion high-level activities as indicated in the -0067 notebook. For ignition frequencies (e.g., plant- instance, fire zone 22 and others are identified in wide values) to estimate the "lead reviews" as additional effort is needed.

physical analysis unit or ignition source level Address the transient ignition sources for frequencies. application fire zones. Providing a complete and through ignition sources identification is important.

IGN-A9-01 POSTULATE the possibility of Dispositioned HBRSEP did postulate transient combustibles Transient fire ignition frequency has been transient combustible fires for for all physical analysis units except for one, assigned to FC490. Transient Influence Factors (CAT I/Il/Ill) all physical analysis units FC490. This physical analysis unit is the of Low(l), Low(l), and Low(l) replaced factors regardless of the administrative deepwell pump D enclosure. HBRSEP needs to No(O), No(O), and No(0) for maintenance, restrictions, provide justification for why transient occupancy, and storage, since entry to FC490, combustibles are not postulated for FC490. which requires using a crane to remove the concrete enclosure, is not precluded.

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SR Topic Status Finding Disposition PP-B6 (01) ENSURE Dispositioned Underground cabling was not addressed in The cable routing database system FSSPMD RNP-F/PSA-0067. The HBRSEP project team was reviewed determining which manholes had (CAT 1/11/111) a. that collectively, the noted a comment was included in this calculation cables within the scope of the Fire PRA defined physical analysis that a "Yard" fire zone should be included in the associated with them. The manholes where units encompass all next revision of the report. cables were identified have been included as locations within the global plant partitioning elements (i.e., Physical Analysis analysis boundary Units) in the Fire PRAs. The following manholes and have been included as plant partitioning elements: MH M-34, MH M-35, MH-1, and

b. that defined physical MH-34. The cables that were identified as routed analysis units do not through MH-35, MH-36 and MH have been overlap included in the Intake Structure Fire Compartment, FC290. The cable loading for the manholes have been assessed following the same approach documented in the HBRSEP combustible loading calc. The approach consists of multiplying the factor of 5,515 BTU/In-ft to each linear ft of cable. Under this approach, a total of 496350 BTUs have been estimated for each manhole. This assumes nine 10' long exposed cables per manhole. Nine cables is the average number of cables identified in the manholes.

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SR Topic Status Finding Disposition PRM-B1 1-01 MODEL all operator actions Dispositioned Main control room abandonment is discussed in Fire response and Main Control Room and operator influences in Section 7.1.4 of Calculation P2217-1022-01-01, Abandonment HFEs have not been incorporated (NOT MET) accordance with the HRA "Fire Human Reliability Analysis" and Table 7.1 in the combined Internal Events/Fire PRA. The element of this Standard. lists the human failure events (HFEs) associated Main Control Room Abandonment HFEs from

[Associated with main control room abandonment. However, Hughes Calculation No.

with HRA-B2 these HFEs have not yet been incorporated into RSC-CALKNX-2013-0301 has been incorporated and PRM-B2] the HBRSEP fire PRA model. Without the HFEs into the recovery rule files used during the for main control room abandonment, the quantification process as appropriate based on HBRSEP fire PRA is incomplete, the fire compartment being quantified.

Incorporate the main control room abandonment RESOLUTION OF CAPABILITY CATEGORY HFEs into the fire PRA model. CLASSIFICATION:

Based on the abovementioned details, this SR is considered to be MET at CAT 1/11/111.

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Duke Energy Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition PRM-B15-01 MODEL any new accident Dispositioned Quantification of the HBRSEP fire PRA model Following additional refinements to the model, progressions beyond the onset indicates that the fire-related LERF is about LERF is more in line with typical results.

(CAT 1/11/111) of core damage identified per 90 percent of the fire-related CDF. This result is PRM-B13 to determine the exceptionally high and has been self-identified fire-induced LERF in by the utility. The result is produced by MSO accordance with HLR-LE-A, events in the switchgear and cable spreading HLR-LE-B, HLR-LE-C, and rooms that cause a core damaging accident HLR-LE-D and their SRs in sequence and also result in the spurious Part 2 with the following opening of several containment isolation valves.

clarifications: Fire-related LERF for HBRSEP is an unusually (a) All the SRs under high proportion of the fire-related CDF. Typically, HLR-LE-A, HLR-LE-B, fire-related LERF is 10 - 25 percent of the HLR-LE-C, and HLR-LE-D in fire-related CDF. At HBRSEP, fire-related LERF Part 2 are to be addressed in is about 90 percent of the fire-related CDF.

the context of fire scenarios including effects on system Investigate ways to remove conservatisms from operability / functionality, the HBRSEP fire PRA model, particularly for the operator actions, accident MSOs in the switchgear and cable spreading progression, and possible rooms.

containment failures accounting for fire damage to equipment and associated cabling.

(b) LE-C2 and LE-C6 in Part 2 are to be met in a manner consistent with 4-2.10.

(c) LE-C6 in Part 2 is to be met in a manner consistent with PRM-B9 above.

(d) LE-C8 in Part 2 is to be met in a manner consistent with PRM-B6 above.

and DEVELOP a defined basis to support the claim of nonapplicability of any of these requirements in Part 2.

Page V-26 HBRSEP LAIR HBRSEP Rev 00 LAR Rev Page V-26

Duke Eneray Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)

SR Topic Status Finding Disposition PRM-Cl-02 DOCUMENT the Fire PRA Dispositioned Documentation of development of the plant The development.of the plant response model is plant response model in a response model is spread over several documented in Attachment 9 of RNP-/PSA-0066 (CAT 1/11/111) manner consistent with calculations and documents. The primary via a model change log. RNP-F/PSA-0066 also HLR-IE-D, HLRAS-C, documents are RNP-F/PSA-0066 (component discusses assumptions made during the HLR-SC-C, HLR-SY-C, and selection) and P2217-1021-01-01 HLR-DA-E and their SRs in (quantification). Thus, key elements of the PRM development of the model.

Part 2 as well as 4-2.10 with development (e.g., assumptions that are made the following clarifications: while developing the model) being dispersed and (a) HLR-IE-D in Part 2 is to be difficult to assimilate and comprehend. The met in a manner consistent dispersion of pertinent information about with that required under development of the PRM makes understanding HLR-IGN-B of this Standard. of the development difficult for reviewers and for (b) Document any defined utility staff who will make future modifications to bases to support the claim of the HBRSEP fire PRA model and nonapplicability of any of the documentation.

referenced requirements in Part 2 beyond that already Develop a PRM-specific calculation or notebook covered by the clarifications in which combines the pertinent portions of PRM this section. documentation from existing documents or calculations.

UNC-A2 INCLUDE the treatment of Dispositioned To comply with Section 2 applicable SRs intent Section 7.0 of the HBRSEP Fire PRA uncertainties, including their for UNC, a final documentation of importance Quantification Calculation, provides an analysis (CAT 1/11/111) documentation, as called out in rankings considering various sensitivity studies of uncertainty regarding the quantification of the SRs PRM-A4, FQ-F1, along with applicable sequences is suggested to HBRSEP Fire PRA. Importance rankings are also IGN-A10, IGN-B5, FSS-E3, be included in the UNC discussion in an discussed in this section. Additionally, FSS-E4, FSS-H5, FSS-H9, and independent document. Section 9.0 provides a discussion of sensitivities CF-A2 and that required by that were evaluated.

performing Part 2 referenced requirements throughout this Standard.

Page V-27 HBRSEP LAR Rev 00 LAR Rev Page V-27

United States Nuclear Regulatory Commission to Serial: RNP-RA/13-0090 2 Pages with Cover Page ENCLOSURE 2 HBRSEP UNIT NO. 2 LIST OF REGULATORY COMMITMENTS

United States Nuclear Regulatory Commission to Serial: RNP-RA/13-0090 Page 2 of 2 H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2 License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Generating Plants (2001 Edition)

HBRSEP LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to in this letter for HBRSEP Unit No. 2.

Any other actions discussed in this submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.

REGULATORY COMMITMENTS DATE Prior to startup after HBRSEP Unit No. 2 will complete those modifications and completion of HBRSEP Unit No. 2 Refueling Outage 31, implementation items outlined in Attachment 'S' to this letter. currently scheduled to begin in the Spring of 2018.

United States Nuclear Regulatory Commission to Serial: RNP-RA!13-0090 2 Pages with Cover Page ENCLOSURE 3 TECHNICAL SPECIFICATIONS PAGES (MARK-UP)

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating procedures required to implement the commitments to NUREG-0737 and of NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; DELETE

• c. Quality assurance for effluent and environmental monitoring;

d. Fire Protectie, Pogram implementationand V e All programs specified in Specification 5.5.

HBRSEP Unit No. 2 5.0-6 Amendment No. 17-

United States Nuclear Regulatory Commission to Serial: RNP-RA/13-0090 2 Pages with Cover Page ENCLOSURE 4 TECHNICAL SPECIFICATIONS PAGES (RETYPED)

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. The emergency operating procedures required to implement the commitments to NUREG-0737 and of NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Quality assurance for effluent and environmental monitoring; and

d. All programs specified in Specification 5.5.

HBRSEP Unit No. 2 5.0-6 Amendment No.