ML11276A120

Enclosure 1 to LIC-11-0099, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

ML11276A120
Person / Time
Site:	Fort Calhoun
Issue date:	09/28/2011
From:	Omaha Public Power District
To:	Office of Nuclear Reactor Regulation
References
LIC-11-0099
Download: ML11276A120 (146)

Text

LIC-1 1-0099 Enclosure 1 Page 1 Omaha Public Power District Allllm -lmill -ENWE,,'E mlllll Fort Calhoun Station, Unit No. 1 Docket No. 50-285/Renewed Facility Operating License No. DPR-40 License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)Transition Report Note: Attachments A, C, D, G, K, S, T, and W contain sensitive security-related information to be withheld from public disclosure pursuant to 10 CFR 2.390.This Enclosure 1 of LIC-11-0099 is decontrolled upon removal of Attachments A, C, D, G, K, S, T, and W.Attachments A, C, D, G, K, S, T, and W of this Enclosure contain SECURITY-RELATED INFORMATION

-WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390.Upon removal of Attachments A, C, D, G, K, S, T, and W, Enclosure 1 is Decontrolled.

Omaha Public Power District FCS NFPA 805 Transition Report Table of Contents List of Acronyms .....................................................................................................

iv Executive S um m ary ...................................................................................................

viii

1.0 INTRODUCTION

..............................................................................................

I 1.1 B a ckg ro u nd ..........................................................................................

..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 rp o se ..............................................................................................

..3 2.0 OVERVIEW OF THE 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 u nd ..........................................................................................

..8 3.2 N FPA 805 Process .................................................................................

8 3.3 NEI 04-02 -NFPA 805 Transition Process ...........................................

10 3.4 NFPA 805 Frequently Asked Questions (FAQs) ...................................

11 4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS

.......................................

12 4.1 Fundamental Fire Protection Program and Design Elements ..............

12 4.1.1 Overview of Evaluation Process ................................................

12 4.1.2 Results of the Evaluation Process .............................................

15 4.1.3 Definition of Power Block and Plant ...........................................

16 4.2 Nuclear Safety Performance Criteria ....................................................

16 4.2.1 Nuclear Safety Capability Assessment Methodology

................

16 4.2.2 Existing Engineering Equivalency Evaluation Transition

...........

25 4.2.3 Licensing Action Transition

......................................................

26 4.2.4 Fire A rea Transition

..................................................................

28 4.3 Non-Power Operational Modes ...........................................................

31 4.3.1 Overview of Evaluation Process ................................................

31 4.3.2 Results of the Evaluation Process ..............................................

35 4.4 Radioactive Release Performance Criteria .........................................

36 4.4.1 Overview of Evaluation Process ................................................

36 4.4.2 Results of the Evaluation Process ............................................

37 4.5 FPRA and Performance Based Approaches

....................

37 4.5.1 FPRA Development and Assessment

.......................................

37 Page i Omaha Public Power District FCS NFPA 805 Transition Report Table of Contents (continued) 4.5.2 Performance-Based Approaches

.............................................

40 4.6 M onitoring Program ............................................................................

44 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring P ro g ra m .................................................................................

..4 5 4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ........ 45 4.7 Program Documentation, Configuration Control, and Quality Assurance 51 4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of N F P A 8 0 5 ..............................................................................

..5 1 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 of N FPA 805 ....................................................................

53 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 8 0 5 ..........................................................................................

..5 7 4.8 S um m ary of R esults .............................................................................

58 4.8.1 Results of the Fire Area Review ................................................

58 4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase ............................................................

59 4.8.3 Supplemental Information

-Other Licensee Specific Issues ........ 60

5.0 REGULATORY EVALUATION

.....................................................................

73 5.1 Introduction

-10 CFR 50.48 ...............................................................

73 5.2 R egulatory Topics ...............................................................................

78 5.2.1 License Condition Changes ......................................................

78 5.2.2 Technical Specifications

...........................................................

78 5.2.3 Orders and Exemptions

............................................................

78 5.3 Regulatory Evaluations

........................................................................

78 5.3.1 No Significant Hazards Consideration

.......................................

78 5.3.2 Environmental Consideration

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79 5.4 Transition Implementation Schedule ....................................................

79

6.0 REFERENCES

.............................................................................................

80 Page ii Omaha Public Power District FCS NFPA 805 Transition Report Table of Contents (continued)

A ttachm ents ................................................................................................................

96 A. NEI 04-02 Table B-1 -Transition of Fundamental FPP and Design E le m e n t ..............................................................................................

..A -1 B. NEI 04-02 Table B-2 -Nuclear Safety Capability Assessment

-M ethodology R eview .....................................................................................

B-1 C. NEI 04-02 Table B-3 -Fire Area Transition

.................................................

C-1 D. NEI 04-02 Non-Power Operational Modes Transition

...................................

D-1 E. NEI 04-02 Radioactive Release Transition

..............

......................................

E-1 F. Fire Induced Multiple Spurious Operations Resolution

..................................

F-1 G. Recovery Actions Transition

........................................................................

G-1 H. NFPA 805 Frequently Asked Question Summary Table ................................

H-1 1. D efinition of Pow er Block ...............................................................................

I-1 J. Fire M odeling 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 C hanges ..........................................................................

M -1 N. Technical Specification Changes ...................................................................

N-1 0 .O rders and Exem ptions ...............................................................................

0 -1 P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)

....................................

P-1 Q. No Significant Hazards Evaluations

.............................................................

Q-1 R. Environmental Considerations Evaluation

....................................................

R-1 S. Plant Modifications and Items to be Completed During Implementation

..... S-1 T. Clarification of Prior NRC Approvals

.............................................................

T-1 U .Internal Events PRA Q uality ........................................................................

U-1 V .Fire P R A Q uality ........................................................................................

..V -1 W .Fire P R A Insights ........................................................................................

..W -1 Page iii Omaha Public Power District FCS NFPA 805 Transition Report List of Acronyms AB Auxiliary Boiler AC alternating current ADAMS Agency-wide Document Access and Management System ADV Atmospheric Dump Valve AFAS Auxiliary Feedwater Actuation Signal AFW Auxiliary Feedwater AHJ Authority Having Jurisdiction Al Auxiliary Instruments ANS American Nuclear Society ANSI American National Standards Institute AON Auxiliary Operator Nuclear AOP Abnormal Operating Procedure ARP Alarm Response Procedure AS Auxiliary Steam ASP Auxiliary Shutdown Panel BAST Boric Acid Storage Tank BOP Balance of Plant BTP branch technical position CCDP conditional core damage probability CCW Component Cooling Water CDF core damage frequency CFAST Collaborative Force Analysis, Sustainment, and Transportation CFR Code of Federal Regulations CIAS Containment Isolation Actuation Signal CIV Containment Isolation Valve.CLB current licensing basis CPHS Containment Pressure High Signal CRS control room supervisor CS Containment Spray CSAS Containment Spray Actuation Signal CST Condensate Storage Tank CVCS Chemical and Volume Control System CW Circulating Water DC direct current DG diesel generator DID defense-in-depth DW Demineralized Water EA engineering analysis ECCS Emergency Core Cooling System ECN Engineering Change Notice EEEE Existing Engineering Equivalency Evaluation EFWST Emergency Feedwater Storage Tank EOP Emergency Operating Procedure EPG Emergency Procedure Guidelines EPIP Emergency Plan Implementing Procedure(s)

Page iv Omaha Public Power District FCS NFPA 805 Transition Report List of Acronyms (continued)

EPRI Electric Power Research Institute EPU Extended Power Uprate ERF Emergency Response Facility ERFBS electrical raceway fire barrier system EROP Equipment Reliability Optimization Program ESC Engineered Safeguards Controls (Normal Operation)

ESF Engineered Safety Feature ESFS Engineered Safety Features System F&Os Facts and Observations FACTS Ft. Calhoun Automated Cable Tracking System FAQ Frequently Asked Question FCS Fort Calhoun Station FCV Flow Control Valve FHA fire hazards analysis FP Fire Protection FPRA fire probabilistic risk assessment FR Federal Register FRE fire risk evaluation FSA fire safety analyses FSAR Final Safety Analysis Report FW Feedwater GDC general design criterion/criteria GL generic letter HCV Hand Control Valve HEAF high energy arcing fault HEP human error probability HGL hot gas layer HRA human reliability analysis HRE higher risk evolution HRR heat release rate HVAC heating, ventilation, and air conditioning I&C Instrumentation and Control IN information notice IPS Iowa Public Service KSF key safety function LAR License Amendment Request LER Licensee Event Report LERF large early release frequency MCB main control board MCC Motor Control Center MCR main control room MCV Motor Controlled Valve MFW Main Feedwater Page v Omaha Public Power District FCS NFPA 805 Transition Report List of Acronyms (continued)

MOU Memorandum of Understanding MOV Motor Operated Valve MS Main Steam MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MSO Multiple Spurious Operation MSSRV Main Steam Safety Relief Valve NEI Nuclear Energy Institute NFPA National Fire Protection Association NPO non-power operational (mode)NRC U.S. Nuclear Regulatory Commission NSCA nuclear safety capability analysis NSPC nuclear safety performance criteria OL Operating License OMA Operator Manual Action ONS Oconee Nuclear Station OP Operating Procedure OPLS Offsite Power Low Signal OPPD Omaha Public Power District OSGSF Original Steam Generator Storage Facility PB performance-based P1 Pressure Indicator PORV Power Operated Relief Valve POS plant operational state PRA Probabilistic Risk Assessment PRC Plant Review Committee PSA probabilistic safety assessment PWR pressurized water reactor QA Quality Assurance QC Quality Control RA recovery action RAI request for additional information RAS Recirculation Actuation Signal RC Reactor Coolant RCP Reactor Coolant Pump RCS Reactor Coolant System RG Regulatory Guide RI risk-informed RI-PB Risk-Informed, Performance-Based RIS regulatory issue/information summary RPS Reactor Protection System RVLMS Reactor Vessel Level Monitoring System RW raw water Page vi Omaha Public Power District FCS NFPA 805 Transition Report List of Acronyms (continued)

RW radioactive waste processing building RWST refueling water storage tank S/G Steam Generator SBO Station Blackout SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent Fuel Pool SFPC Spent Fuel Pool Cooling SFPE Society of Fire Protection Engineers SG steam generator SI Safety Injection SIAS Safety Injection Actuation Signal SIRWT Safety Injection and Refueling Water Tank SM Shift Manager SOP System Operating Procedure SOV Solenoid Operated Valve SRO Senior Reactor Operator SSC system, structure, and component SSD Safe Shutdown SSE Safe Shutdown Equipment SSEL Safe Shutdown Equipment List STA Shift Technical Advisor SW Service Water (potable water)T-H thermal-hydraulic TPCW Turbine Plant Cooling Water TR Temperature Recorder TR Transition Report TS Technical Specifications TSC Technical Support Center UFHA Updated Fire Hazards Analysis USAR Updated Safety Analysis Report V&V Verification and Validation VCT Volume Control Tank VFDR variance from deterministic requirements ZOI zone of influence 4kV 4.16kV or 4160 VAC electrical distribution system (used interchangeably)

Page vii Omaha Public Power District FCS NFPA 805 Transition Report Executive Summary The Omaha Public Power District (OPPD) will transition the Fort Calhoun Station (FCS), Unit No. 1 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 10 CFR 50.48(a), 10 CFR 50.48(b), and 10 CFR 50, Appendix R, will be superseded.

OPPD submitted the letter of intent to adopt NFPA 805 in accordance with 10 CFR 50.48(c) on June 9, 2008. FCS began the transition to the performance-based standard for fire protection in June 2008.The transition process consisted of a review and update of FCS documentation, including the development of a Fire Probabilistic Risk Assessment (PRA) using NUREG/CR 6850 as guidance.

This Transition Report (TR) summarizes the transition process and results. This TR 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 (RG) 1.205 Revision 1.Section 4 of the TR 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: " Non-Power Operational Modes" Fire Risk Evaluations" Radioactive Release Performance Criteria" Monitoring Program* Program Documentation, Configuration Control, and Quality Assurance Section 5 of the TR 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 TR 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 Page viii Omaha Public Power District FCS NFPA 805 Transition Report 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this License Amendment Request (LAR). The methodologies associated with these FAQs have been included in the TR for Nuclear Regulatory Commission (NRC) approval.Page ix Omaha Public Power District FCS NFPA 805 Transition Report

1.0 INTRODUCTION

The 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). Omaha Public Power District (OPPD) is implementing the Nuclear Energy Institute methodology NEI 04-02, Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program under 10 CFR 50.48(c), to transition FCS 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 FCS 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, 10CFR50.48(c), which establishes new Risk-Informed, Performance-Based (RI-PB) fire protection requirements.

10CFR50.48(c) incorporates by reference, with exceptions, 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), for 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: 'Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.Page 1 Omaha Public Power District FCS NFPA 805 Transition Report ES£r I Incorporation by.....__. Reference Qbnd-ed for F1P for tight Water Reactor Eleotri' Generating P1311113 10 CFR..... 50.48(c)National Fire Protection Association Standard NFPA 805."0"5 C, NEI 04-02 GUIDANCE FOR IMPLEMENTING A RI-PB FP PROGRAM UNDER 10 CFR 50.48(c)Endorsement RG 1.205 RI-PB FP FOR EXISTING LIGHT-WATER NUCLEAR POWER PLANTS Figure 1-1 -NFPA 805 Transition

-Implementation Requirements/Guidance

1.1.2 Transition

to 10 CFR 50.48(c)1.1.2.1 Start of Transition OPPD submitted a letter of intent to the NRC on June 9, 2008 (ML081620232) for FCS to adopt NFPA 805 in accordance with 10 CFR 50.48(c).On September 10, 2008, the NRC published in the Federal Register (73 FR 52705) a revision to its Interim Enforcement Policy regarding enforcement discretion for certain fire protection issues, allowing licensees the option to request an extended enforcement discretion period for submittal of a LAR if they are pursuing transition to 10 CFR 50.48(c).

This revision states that an additional period of enforcement discretion may be granted on a case-by-case basis, if a licensee has made substantial progress in its transition effort. This additional period of enforcement discretion, if granted, would end six months after the date of the safety evaluation approving the second pilot plant LAR review. The enforcement discretion will continue in place, without interruption, until NRC approval of the LAR to transition to 10 CFR 50.48(c).By letter dated November 20, 2008 (ML082710002), the NRC responded to OPPD's letter of intent. The letter stated that the NRC considers the discretion period for FCS to begin on June 9, 2008, and to expire on June 9, 2011. The NRC letter identified that the NRC had granted a third year of enforcement discretion by Federal Register Notice 71 FR 19905 dated April 18, 2006. In accordance with NRC Enforcement Policy, the Page 2 Omaha Public Power District FCS NFPA 805 Transition Report enforcement discretion period will continue until the NRC approval of the LAR is completed.

By letter dated July 1, 2009, the NRC informed NEI of a change in the review schedule of the second pilot plant adopting NFPA 805 and stated that plants that began transition before September 30, 2007, may apply for extensions to their enforcement discretion period under the enforcement policy described in 73 FR 52705. However, plants that began transition after September 30, 2007, which includes FCS, must maintain their current submittal schedule.1.1.2.2 Transition Process The transition to NFPA 805 includes the following high level activities: " Complete safe shutdown analysis/10 CFR 50, Appendix R Reconstitution;" Develop Fire PRAs using NUREG/CR 6850 as guidance and revise Internal Events PRA to support the Fire PRAs; and,* Completion of activities required to transition the pre-transition Licensing Basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205.1.2 Purpose The purpose of the TR is as follows: 1. Describe the process implemented to transition the current fire protection program to compliance with the additional requirements of 10 CFR 50.48(c);2. Summarize the results of the transition process;3. Explain the bases for conclusions that the fire protection program complies with 10 CFR 50.48(c) requirements;

4. Describe the new fire protection licensing basis, and 5. Describe the configuration management processes used to manage post-transition changes to the station and the Fire Protection Program, and resulting impact on the Licensing Basis.Page 3 Omaha Public Power District FCS NFPA 805 Transition Report 2.0 OVERVIEW OF THE EXISTING FIRE PROTECTION PROGRAM 2.1 Current Fire Protection Licensing Basis FCS Unit No. 1 was licensed to operate on August 9, 1973. As a result, the FCS fire protection program is based on compliance with 10 CFR50.48(a) and (b) and 10 CFR 50 Appendix R and the following License Condition:

FCS Operating License Condition 3.D for Unit 1 states: D. Fire Protection Program Omaha Public Power District shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report for the facility and as approved in the NRC safety evaluation reports (SERs) dated February 14 and August 23, 1978; November 17, 1980; April 8 and August 12, 1982; July 3 and November 5, 1985; July 1, 1986; December 20, 1988; November 14, 1990; March 17, 1993; and January 14, 1994, subject to the following provision:

Omaha Public Power District 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.On November 4, 2003, Renewed Facility Operating License No. DPR-40 was issued for FCS.2.2 NRC Acceptance of the Fire Protection Licensing Basis The Current Licensing Basis (CLB) for the Fire Protection (FP) Program at FCS consists of Title 10 Code of Federal Regulations Part 50 (10 CFR 50) Section 50.48, 10 CFR 50 Appendix A General Design Criteria (GDC) 3 draft criteria, 10 CFR 50 Appendix R Sections lIl.G, II.J and 111.0, Appendix A to Nuclear Regulatory Commission (NRC)Branch Technical Position (BTP) APCSB 9.5-1, Renewed Facility Operating License (OL) No. DPR-40, including the Technical Specifications (TS), the Updated Safety Analysis Report (USAR), the Quality Assurance (QA) Plan and various NRC Safety Evaluation Reports (SERs) that have approved implementation of the FP Program and, as appropriate, variations from regulatory requirements.

Most of these documents did not exist at the time FCS received its OL. FCS was licensed in accordance with the 70 draft General Design Criteria (GDC) published for comment in the Federal Register (32 FR 10213) on July 11, 1967. 10 CRF 50, Appendix A contains updated (non-draft)

NRC GDC 3. The FCS USAR Appendix G criterion 3 satisfies the 10 CFR 50, Appendix A, Criterion

3. The description of the FP system is provided in the FCS Updated Safety Analysis Report (USAR). The majority of FP regulations and subsequent implementing documents were developed in response to a fire at the Browns Ferry Unit 1 nuclear plant on March 22, 1975.Page 4 Omaha Public Power District FCS NFPA 805 Transition Report Subsequently, FCS was required to perform an analysis of its FP Program relative to the requirements of Appendix A to BTP APCSB 9.5-1, which had been developed to identify minimum requirements for FP Programs at nuclear plants docketed prior to July 1, 1976. The results of this analysis, including proposed modifications to the facility and changes to the FP Program, were submitted to the NRC in letter LIC-76-0180, dated December 30, 1976. FCS was also provided with model TS associated with FP, and was requested to develop and submit proposed FP TS for FCS. The initial submittal of FP TS for FCS was made in letter LIC-77-0172, dated February 16, 1977. The first approved version of FP TS was documented in a NRC SER dated February 14, 1978.(Amendment No. 38) An NRC SER dated August 23, 1978 (Amendment No. 40), approved a revision to the FP TS and documented assessment of the FCS FP Program.This SER added a license condition to track the completion of 35 required modifications to the facility or changes to the Program. It was also noted, "...there remains certain items, principally information needed to demonstrate capability for safe shutdown in the event of certain fires, which must be completed before we can reach a determination concerning whether the fire protection program at this facility will satisfy the objectives of Appendix A of BTP 9.5-1, for satisfactory long term fire protection." An NRC SER dated November 17, 1980, approved a revision to the FP TS and documented progress on open items. (Amendment No. 53)On February 17, 1981, 10 CFR 50.48 became effective.

This regulation requires each nuclear plant to maintain a fire protection plan that satisfies GDC 3 of Appendix A to 10 CFR 50, and requires all nuclear plants licensed to operate prior to January 1, 1979 to meet the requirements of Sections Ill.G, IIl.J and 111.0 of Appendix R to 10 CFR 50.From this point forward, FCS analyses associated with safe shutdown and alternate shutdown in the event of a fire were required to comply with Sections III.G and III.L (referenced in III.G for alternate shutdown) of Appendix R. Likewise, emergency lighting was required to comply with Section III.J and an oil collection system that complied with Section 111.0 was required for reactor coolant pumps (RCPs).An NRC SER dated April 8, 1982, determined that the alternate shutdown design for a fire in the Control Room or Cable Spreading Room met the requirements of III.G and lII.L of Appendix R with three (3) exceptions.

The SER required installation of a source range flux monitor on the alternate shutdown panel, revising the design to consider the failure of non-safety related associated circuits which could affect safe shutdown by shorts or grounds, and modifying the plant such that achieving and maintaining hot standby could be accomplished without pulling fuses or making wiring changes. A subsequent NRC SER dated August 12, 1982, documented that FCS had committed to make the appropriate changes and, based on that, determined that the FCS design met the requirements of III.G and III.L of Appendix R.During the week of May 16, 1983, the NRC conducted an inspection at FCS of the implementation of, and compliance with, the safe shutdown requirements of Section III.G of Appendix R. This inspection identified four (4) violations, some with multiple examples.

As a result, on August 30, 1983, FCS submitted requests for exemption from the requirements of Section III.G of Appendix R that were associated with the identified violations.

Following several discussions with the NRC staff, and additional analysis, on December 3, 1984 FCS submitted an additional exemption request for the Containment Page 5 Omaha Public Power District FCS NFPA 805 Transition Report Building.

In addition, on January 9, 1985, FCS submitted a request to delete one of the exemptions that had been requested on August 30, 1983 and provided revised justification for three (3) others. In an SER dated July 3, 1985, the NRC approved exemptions from Section III.G.2 of Appendix R for Containment, the Intake Structure and pull boxes on the exterior south wall of the Auxiliary Building, Room 19, Room 20 and Room 56; and approved exemptions from Section III.G.3 of Appendix R for Room 57W and the Control Room.An NRC SER dated November 5, 1985, approved the alternate shutdown methodology for a fire in Room 57W. Clarifications/Changes to the July 3, 1985, SER were approved in an NRC SER dated July 1, 1986. Specifically, this SER approved changes to the separation of redundant equipment in Containment, elimination of a shield intended to protect the motor-driven auxiliary feedwater pump from water impingement, and lack of sprinkler coverage in Corridor 53, and clarified the location of in-tray sprinkler heads in Room 19. This SER also, incorrectly, clarified that a one-hour rated fire wrap was not required on specific cables in Room 20 because the cables in question had been rerouted outside of the fire area. The cables were, in fact, maintained in Room 20 and appropriately wrapped. This was subsequently corrected in an NRC SER dated March 17, 1993.An NRC SER dated December 20, 1988, approved an exemption from Section 111.0 of Appendix R for undersized collection tanks associated with the RCP lube oil collection system. An exemption request associated with inadequate separation of redundant safe shutdown cabling in Room 57W was denied in an NRC SER dated November 14, 1990.Amendment No. 160 to the FCS OL, and its associated SER, were issued on January 14, 1994. This amendment relocated FP TS, with some approved modifications, to the USAR in accordance with NRC Generic Letters (GL) 86-10 and 88-12. It also added the following FP License Condition, which includes each of the SERs described previously:

Omaha Public Power District shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report for the facility and as approved in the SERs dated February 14, and August 23, 1978, November 17, 1980, April 8, and August 12, 1982, July 3, and November 5, 1985, July 1, 1986, December 20, 1988, November 14, 1990, March 17, 1993, and January 14, 1994, subject to the following provision:

Omaha Public Power District 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.In addition to the SERs specifically referenced in the FP License Condition, two (2)other SERs document significant licensing bases related to FP. An NRC SER dated December 22, 1986, approved the use of carpet with a Critical Radiant Heat Flux of 0.45 watts/cm2, per ASTM E-648, as equivalent to carpet with Flame Spread, Smoke and Fuel Contribution ratings of 25 or less, per ASTM E-84, as specified in Appendix A of BTP APCSB 9.5-1, Section D.1(d). An NRC SER dated May 21, 1998, approved an Page 6 Omaha Public Power District FCS NFPA 805 Transition Report exemption from Section 111.0 of Appendix R for five (5) unpressurized locations on the ABB RCP motor. It also determined that exemptions were not required for lack of a flame arrestor on the RCP oil collection tanks and not having oil collectors at the anti-rotation device housing and the motor cooling air vents on GE RCP motors.An NRC letter dated February 6, 2009, approved and transmitted an exemption from specific requirements of Section IIl.G.1.b of Appendix R. Specifically, the exemption addresses the power and control cables for the four raw water pumps that are routed from the auxiliary building through outside cable pull boxes 128T and 129T into the underground duct bank and manhole vault numbers 5 and 31 into the intake structure building.

As discussed above, by letter dated July 3, 1985, the NRC granted an exemption from the technical requirements of Section III.G.2 of Appendix R to 10 CFR Part 50, for Fire Area 31 (intake structure building) and for the pull box area of the auxiliary building.

However, cables in the duct bank and manhole vault numbers 5 and 31 that are routed between the pull boxes and intake structure were not discussed in the OPPD August 30, 1983, exemption request. The OPPD letter dated February 4, 2008, clarified that the NRC July 3, 1985, SER incorrectly referenced Section III.G.2 and subsequently provided exemption from 10 CFR 50, Section IIl.G.l.b.

Specifically, the OPPD letter of February 4, 2008, states that "This exemption request thereby provides notification and clarification that the original SER and exemption should have referenced 10 CFR 50, Appendix R, Section III.G.1.b

....." Therefore, the NRC staff evaluation in letter dated February 6, 2009 supersedes the reference "Section III.G.2" used in the NRC SER dated July 3, 1985, for Fire Area 31 (intake structure building and pull boxes 128T and 129T outside the auxiliary building).

OPPD identified the duct bank and manhole vault numbers 5 and 31 as part of the Fire Area 31; therefore, the NRC staff evaluation in the February 6, 2009, letter approved exemption from 10 CFR Part 50, Appendix R, Section IIl.G.1.b for duct bank and manhole vault numbers 5 and 31.NRC inspection report 05000285/2011003 dated August 11, 2011, identified a noncited violation of 10 CFR 50, Appendix R, Section 111.0 for the failure to ensure an adequate seismic design of the RCP lube oil collection system. This was entered into the FCS corrective action program under condition report 2011-6631 for resolution.

Page 7 Omaha Public Power District FCS NFPA 805 Transition Report 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, FCS has implemented the NFPA 805 Section 2.2 process by first determining the extent to which its current fire protection program supports findings of deterministic compliance with the requirements in NFPA 805. RI-PB methods are being applied to the requirements for which deterministic compliance could not be shown.Page 8 Omaha Public Power District FCS NFPA 805 Transition Report Estabehf fundanwiflat I- NFPA WS5 Secticn 2.2112)rOldrecOin eleneleb (Chapter 3)tdete hw "& haads FPA SOS Section 2.2(b)WIPA 8065 ecthoo 2.2(c)Evaluate comptiantpe to perforrmane ciritewia Nuceaw safety iitdsnefy pelorrniace critefa to be Lee safety ejcwwined Prop"rt 8wagnmerAusnes (Chapter 1) tritenruptiof 4! k < PAXUon release klenttty ssuntsxe, syrsterms o corivotmi5 (SSC5ISt in eac NFPA 06 Section 2.2"8 W~asto uhich the peifoirriwce crtrau applies-4-DOWtlimsinac Approach Meant"~ corngiance With euatm9 plant license basis 10 CFR 50 App R. Appgroved Exemnronh.

Engsieeivi Evaluations)

NFPA 806 Section 2-2(e)I, Perfomance-Based'AjtprCh Evaluate Mtbt to satisfy perlomiane rejureemfs (Chapter 4)4-,1, -D~eteministic 8"4 Verily deewitcreqwifenlts wm'el LErg~wg Pe~rfmencri Blaru Ddefw fire scenamoa end fve design be"s for each fre ama batng considered, Evalumt using, e g.*Fire IOdeMM to WW"ynfire Mk and margin f safety*PSA lo examne knpact on overall plri rsk WfPA W0 SecbtMo 2.2M tf'PA $05 Section 2.2(g)Risk-Infonned Change Evaluation WPA WS Eveate risk mpact of chnges to the approved design basts Feedback NFPA 805-4 s-lacingbogr Section 2.2(h)IfPA Design 5asi Documnerns Fre hazards anayis wa ria" cabaiy meme suippomtwn engineering caieuiatons prbabostc saet analyss Risk-lomred change evaktons Figure 3-1 -NFPA 805 Process (NEI 04-02 Figure 3-1 based on NFPA 805 Figure 2.2) 2 2 Note: 10 CFR 50.48(c) does not incorporate by reference Life Safety and Plant DamagelBusiness Interruption goals, objectives and criteria.See 10 CFR 50.48(c) for specific exceptions to the incorporation by reference of NFPA 805, Page 9 Omaha Public Power District FCS NFPA 805 Transition Report 3.3 NEI 04-02 -NFPA 805 Transition Process NFPA 805 contains technical processes and requirements for a risk-informed, performance based 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 one based upon NFPA 805, as shown below in Figure 3-2.Transition Report SecL 4.1 FP Fundamentals Review and Confirmation LI Identify outliers / VFDRs]Transition Report Sect. 4.2 Nuclear Safety Review and Confirmation Identify outliers / VFDRs Transition Report Sect. 4.4 Transition Report Sect. 4.3 I I 1 FP Fundamentals Assessment Perform Engineering Analyses Radioactive Release Assessment Non-power operational mode Assessment Nuclear Safety Analyses Use PB Approach if Needed (Fire Modeling or Fire Risk Evaluations)

}Transition Report Sect. 4.5 Verify / Establish Monitoring Program i Confirm / Establish Adequate Documentation

/ Quality and Configuration Control} Transition Report Sect. 4.6}STrans tion Report Sect. 4.7, 5}STransition Report ect. 4.8, 5 Regulatory Submittal and Approval Figure 3-2 -Transition Process (Simplified)(based on NEI 04-02 Figure 4-1)Page 10 Omaha Public Power District FCS NFPA 805 Transition Report 3.4 NFPA 805 Frequently Asked Questions (FAQs)The NRC has worked with NEI and two Pilot Plants (ONS 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 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 transition 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.OPPD intends to utilize guidance from FAQ 10-0059, Monitoring Program, for the development and implementation of the FCS NFPA 805 monitoring program when the FAQ is approved by the NRC. Development and implementation of the NFPA 805 monitoring program for FCS will be completed as part of LAR implementation. (See Attachment S). OPPD does not regard this as being a deviation from the approved guidance.OPPD has utilized guidance from a draft version of FAQ 08-0050, "Non-Suppression Probability," dated May 30, 2008. This draft version of the FAQ was implemented because it was the most current version when the relevant OPPD task was initiated.

The more recent version of FAQ 08-0050 is documented in NUREG/CR-6850 Supplement 1, issued September 2010. This version has been reviewed, and it is qualitatively judged that implementation of the more current version will not impact the conclusions of this NFPA 805 LAR. This judgment is primarily based on the low frequency of fire events in which the FCS FPRA primarily credits manual suppression (i.e., scenarios leading to hot gas layer formation).

The preparation of this LAR has included the review and incorporation, as applicable, of requests for additional information (RAIs) from the NFPA 805 Pilot Plants, Harris and Oconee.Page 11 Omaha Public Power District FCS NFPA 805 Transition Report 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 FCS Fire Protection Program to the requirements of NFPA 805 Chapter 3 was performed and documented in a FCS Engineering Analysis EA10-062 entitled "NFPA 805 Chapter 3 Fundamental Fire Protection Program and Design Elements Review." EA10-062 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. (NOTE: Where this compliance statement is used, the clarification is provided in the Compliance Basis column (See Attachment A)).* Complies by previous NRC approval -For those sections/subsections where the specific NFPA 805 Chapter 3 requirements are not met but previous NRC approval of the configuration exists." Complies with use of Existing Engineering Equivalency Evaluations (EEEEs) -For those sections/subsections determined to be equivalent to the NFPA 805 Chapter 3 requirements as documented by engineering analysis.* Complies, with Required Action -For those sections/subsections determined to meet the specific requirements of NFPA 805 after the completion of a modification or other implementation item, such as a procedure change or a work request. (See Attachment S for details)." Submit for NRC Approval -For those sections/subsections for which approval is sought in this LAR submittal in accordance with 10 CFR 50.48(c)(2)(vii).

A summary of the bases of acceptability is provided (See Attachments L or T for details as applicable).

Page 12 Omaha Public Power District FCS NFPA 805 Transition Report 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 13 Omaha Public Power District FCS NFPA 805 Transition Report Existing Fundamental Fire Protection1 Program and Design Element EtrIn the Compliance Basis Field In Reference Document Field omelnyOn in Compliance Statement Field -No Addilionall Ctarification

-Document References that (see Note 1) demonstrate compliance meets yes th FA805 Chpter 3 F Requirement

?< ii ý > Enter In the Compliance Basis Field In Reference Document Field-ne I provideeve-Itemsefoundtduring

'Complies with Clartiicatiou n provide: -Dcument Referencesylhal in Compliance Statement Field Details Clarifycation -Demorenate complncest In the Compiiance Basis Field In Reference Document Field NFPA 805 Enter triovide verbratime excerptl from; -viosigD ocmn Dcmn ayOe Catr3Rqieet Yes 'Compines yPeviogEus'-c -C Appoa doumn Lienin pv D ocmn DcmnayOe hes previous Approva- Subreityaosuent d ou ttem found during approval?

in Compliance Statement Field necessary to cLante Approumen Rfect approval)Demonrstrate Compliance No Ene In thepliance BestFel ~ poie Existing Yes pie Enthuer on thetin Cni anc ai Field In Reference Document FieldDouetayOn gineering uvaten y e Compsiewhe atis Reqie: prcvide: Ioctues aond drng a.o Engineering Eqr ..atency -1 pr-vd I t d EelNot2 Evaluations (BEEcE) Summery of subject of -ce Document R ences thatReve oein Comliance Statement Field Engineering Evaluaion SEnter3'Further Action 1 h Reference Document NRCnmpinae A isro Required in Compliance In Compliance Basis Field Field provide:-W -- SAlpiotateF-ri

_ta -vi -11* Corrective Actions as -Acti ReqredStateerl Field List of Actions to he Taken appropriate Doe net ( Document References Chapter 3 nor is there prevous pproalhr Reference Document Enter 'Sabmt for NRC In Comrpliance Basis Field Field plroide:__NRC Appr~oval -a Approver in Compliance

_1 provde; -11 Cerrectine Actiens. as -Choose On is Required Sataemert Field -Summary of subject of appropriate Requet" for NRC Approva I Document References NOTE 1;Although not included in the submittal to the NRC. the basis for compliance should be recorded in the transition database for reference purposes.NOTE 2: Existing Engineereg Equivalency Evaluations, previously known as Generic Letter 86-.t0 oviluations, exemptions, and deviations were performed for fire protection design variances such as fire protecdon system designs and fire barrier component deviations from the specific fire protection deterministic requremente.

Section 2.2.7 of NFPA805 allows existing BEE that clearly demonstrates an equivalent level of fire protection compared to the deterministic rnqulrenentas to be transitioned.

1hose EEEEs that demonstrate that a system or feature is rated or compliant (i.e, equivalent level of protectioni do not need to be summarized and subimilted as part of the LAR. Those E--EEs against Sections 3.1 through 37 of NFPA 805 that determine the fire protection system or feature is adequate for the hazard should be summa-ized in the LAR.NOTE 3: Al 'further action required' shrold be addressed prior to submittal to the NRC. Actions that remain open wmit be identified as'limplementation Items'that will reference a plant notification number and sill demonstrate that the attribute will be in compliance once the action is closed.104o21 r 21t1 O Document any Open s Items found during Review-Items found during Review In Cempliance Basis Field In Reference Domument Compliance witt be achieved Enter'Complies, with prvmide: Field providei Document any Open-so through a mod orother -w Required Action' In -Summary CoodrAnivedActpo ns, as -D- temsefund duOng action. compliance fiettt lobe taken CreieIesfuddrn Identify Implementation Item appropriate Review L andor Corrective Action Document References I numnber as appropriate Figure 4-1 -Fundamental Fire Protection Program and Design Elements Transition Process (Based on NEI 04-02 Figure 4-2)3 3 Figure 4-1 depicts the process used during the transition and therefore contains elements (i.e., open items) that represent interim resolutions.

Additional detail on the transition of EEEEs is included in Section 4.2.2.Page 14 Omaha Public Power District FCS NFPA 805 Transition Renort 4.1.2 Results of the Evaluation Process 4.1.2.1 NFPA 805 Chapter 3 Requirements Met or Previously Approved by the NRC Attachment A contains the NEI 04-02 Table B-i, Transition of Fundamental FP 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 FCS either:* Complies directly with the requirements of NFPA 805 Chapter 3,* Complies with clarification with the requirements of NFPA 805 Chapter 3,* Complies with 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. FCS requests that the NRC concur with their finding of prior approval for the following sections of NFPA 805 Chapter 3:* 3.3.4 -Clarification is requested for two polyurethane foam monoliths installed in manhole 31.The discussion of the prior approval, including appropriate reference documents, is provided in Attachment T.4.1.2.3 NFPA 805 Chapter 3 Requirements Not Previously Approved by NRC The following sections of NFPA 805 Chapter 3 are not specifically met nor do previous NRC approvals of alternatives exist: " 3.3.1.2(1)

-Approval is requested for the use of commonly available equipment containing non fire-retardant wooden components, such as hand tools.* 3.3.1.2(3)

-Approval is requested for the use of temporary scaffolding that is constructed in place, and the use of combustible equipment/supplies that are not easily relocated due to size, weight, or bulk.* 3.3.5.1 -Approval is requested for the existence of cable (which is not approved for plenum use and not installed in conduit or cable trays) above suspended ceilings in the control room and personnel complex area.Page 15 nvnhn kIJ,jhi Dr1,AIr flicfrirf-f: Q AIJ=IDA JQ09; Trnneifinn Po n* 3.5.3, 3.5.6 (NFPA 20-1996, Section 7-5.2.3) -Approval is requested for the ability to stop the electric fire pump remotely from the control room and from the associated 4160V switchgear.

• 3.5.14 -Approval is requested for the existence of curb valves sectionalizing the underground yard fire main loop that are not supervised, locked, or sealed.* 3.6.1 (NFPA 14-1996, Section 2-7.2) -Approval is requested for the use of lengths of hose at hose stations greater than lengths allowed by NFPA 14.* 3.11.5 -Approval is requested for overhead cabling encased in conduit, wrapped in metal lath, and surrounded by 2 inches of Pyrocrete in fire area 36A.The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements are provided in Attachment L. FCS requests NRC approval of these performance-based methods.4.1.3 Definition of Power Block and Plant Where used in NFPA 805, Chapter 3, the terms "Power Block" and "Plant" refer to structures that have equipment required for nuclear plant operations, such as 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.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 Page 16 Omaha Public Power District FCS NFPA 805 Transition Renort fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3.4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process NFPA 805 Section 2.4.2 Nuclear Safety Capability Assessment states: "The purpose of this section is to define the methodology for performing a nuclear safety capability assessment.

The following steps shall be performed:

1. Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1 2. Selection of cables necessary to achieve the nuclear safety 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 methodology in the following FCS 10 CFR 50 Appendix R and NFPA 805 engineering analyses against the guidance provided in NEI 00-01, Revision 1 Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02." EA-FC-97-044

-10 CFR 50 Appendix R Cable Identification" EA10-037 -Task 7.3 Fort Calhoun Station NFPA 805 NSPC and Fire PRA Circuit Analysis, Cable Selection, and Cable Location" EA10-036 -Task 4.2 Fort Calhoun Station Automation and Update of Safe Shutdown Analysis The methodology is depicted in Figure 4-2 and consisted of the following activities:

• Each specific section of NFPA 805 Section 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: " Aligns" Aligns with intent" Not in Alignment" Not in Alignment, but Prior NRC Approval" Not in Alignment, but no adverse consequences Page 17 Omaha Public Power District FCS NFPA 805 Transition Repo 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. (Note: Comparison of the FCS NSCA methodology to Chapter 3 of NEI 00-01 Revision 1 determined that the methodology does not fail to align with the guidance in any applicable section of NEI 00-01.)The comparison of the FCS deterministic NFPA 805 safe shutdown methodology to NEI 00-01 Chapter 3 (NEI 04-02 Table B-2) was performed and documented in FCS engineering analysis EA10-064, Fort Calhoun Station Nuclear Safety Performance Analysis Methodology Review.Results from Evaluation Process The method used to perform the FCS deterministic NFPA 805 safe shutdown analysis 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 directly or met the intent of the endorsed guidance with adequate justification as documented in Attachment B.Page 18 Omaha Public Power District FCS NFPA 805 Transition Renort Omaha Public Power District FCS NFPA 805 Transition Renort Step I Assemble Docurmentation Determine and Document Step 2 Applicability of NEI 00-01 Sections FI0m GdpeG c al NE I ui-01 Stsp3 Sect i(Conder eny in correne Step 4 Figure 4-2 -Summary of Nuclear Safety Methodology Review Process (FAQ 07-0039)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.Per NFPA 805 the definition of "safe and stable" is (Ref. NFPA 805, definition 1.6.56): Safe and Stable Conditions.

For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff < 0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby (FCS clarification:

the NFPA 805 term "hot standby" for FCS corresponds to Mode 3[Hot Shutdown Condition], per the FCS Technical Specifications) for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff < 0.99 and fuel coolant temperature below boiling (F4S clarification:

the NEPA 805 term "below boiling" for ECS corresponds to "sub-cooled, not boiling at pressure").(Ref. FOS Technical Specifications, Definitions, as of Amendment No. 265, and EOP-20, "Functional Recovery Procedure")

Page 19 Omaha Public Power District FCS NFPA 805 Transition Repo The nuclear safety goal of NFPA 805 requires "...reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition" without a specific reference to a mission time or event coping duration.For the plant to be in a safe and stable condition, it may not be necessary to perform a transition to cold shutdown as currently required under 10 CFR 50, Appendix R.Therefore, the unit may remain at or below the temperature defined by a hot shutdown plant operating state for the event.Results The NFPA 805 Nuclear Safety Performance Criteria (NSPC) Analysis for FCS has been developed to ensure that the plant can achieve and maintain the fuel in a "safe and stable" condition assuming that a fire event occurs during FCS Mode 1 (Power Operation Condition), Mode 2 (Hot Standby Condition), or Mode 3 (Hot Shutdown Condition).

The objective of the deterministic NFPA 805 safe shutdown analysis is, for an all consuming fire occurring in any one plant fire area, to demonstrate that FCS can achieve and maintain Mode 3 (Hot Shutdown Condition), with the minimum plant operating shift staff, for a coping time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The 24-hour coping time has been selected by FCS based on the design capacity for the backup nitrogen supply that is relied upon to maintain positive remote control over the turbine driven auxiliary feedwater pump, and based on the ability of the FCS Emergency Response Organization to respond to the event, with adequate time allowed for the ERO personnel to muster, assess the extent of fire damage, and assist the plant operating staff with implementation of the required actions to sustain Mode 3 (Hot Shutdown Condition), beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or to assess the extent of fire damage, and assist the plant operating staff with implementation of cold shutdown actions and/or cold shutdown repairs for the plant to transition to, and enter, Mode 4 (Cold Shutdown Condition).

Actions required to sustain Mode 3 (Hot Shutdown Condition), beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> include actions to assume local manual control of the turbine driven AFW pump, and provide diesel fuel oil for the emergency diesel generator and/or the diesel driven AFW pump (as necessary, for those fire areas where Offsite Power is not free of fire damage, and/or where the diesel driven AFW pump is credited for NFPA 805 safe shutdown).

The manual operator actions identified below are directed by existing FCS plant operating procedures and may be required within the 24-hour coping time for NFPA 805 safe and stable plant operation.

However, these actions have not been identified in the NFPA 805 safe shutdown analysis as non-compliances or as Variances from Deterministic Requirements (VFDRs) of NFPA 805, as the actions are already part of the existing FCS plant licensing basis, are part of the existing FCS plant design, or are included in the IE PRA and the FPRA.* Refill of the emergency feedwater storage tank (EFWST) per AOP-30,"Emergency Fill of Emergency Feedwater Storage Tank" -approximately 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Page 20 Omaha Public Power District FCS NFPA 805 Transition ReDort actions, FCS has no capability from the main control room to do this automatically or manually." Loss of instrument air, per AOP-17, "Loss of Instrument Air" (specifically, approximately 4-hour actions to isolate flow through condensate makeup control valve, LCV-1190, to prevent CST draindown into the hotwell) for NFPA 805, LCV-1 190 isolation action is only necessary when the CST is used (when the diesel driven AFW pump is credited)." Loss of instrument air, per AOP-17, (specifically, approximately 12-hour actions to prevent a recirculation actuation signal [RAS] from safety injection and refueling water storage tank low level signal [STLS]).The following provides additional detail regarding the "safe and stable" plant operation objective for the FCS deterministic NFPA 805 safe shutdown analysis: The FCS deterministic NFPA 805 safe shutdown analysis includes assessment of the fire impact upon the FCS plant systems and features that are required to achieve and maintain Mode 3 (Hot Shutdown Condition), from Mode 1 (Power Operation Condition).

The deterministic NFPA 805 safe shutdown analysis adequately bounds the objective of"safe and stable" plant operation for a fire event occurring in Mode 2 (Hot Standby Condition), or Mode 3 (Hot Shutdown Condition).

The Mode 3 applicability for the deterministic NFPA 805 safe shutdown analysis is defined as being up to the point at which the MCC breakers for the shutdown cooling suction valves, HCV-347 and HCV-348, are un-locked and closed, at which point spurious operation of these high/low pressure interface valves can occur due to fire damage to the valve control circuitry.

1. FCS has design features and procedures to ensure that an adequate source of inventory is provided for decay heat removal in sustained Mode 3 conditions (i.e., EFWST re-fill capability from raw water for the motor driven AFW pump and the turbine driven AFW pump [Ref. AOP-30, "Emergency Fill of Emergency Feedwater Storage Tank"], and/or alternate water supply for the diesel driven AFW pump from the main condenser upon depletion of the CST).2. Core decay heat in the Mode 3 (Hot Shutdown Condition) will be rejected to the secondary plant through one or both of the steam generators, and then to atmosphere through the main steam safety relief valves operating as spring relief valves.3. The FCS reactor core design ensures that Keff is maintained

<0.99 while the plant is in sustained Mode 3 (Hot Shutdown Condition).

Consequently, maintaining the "safe and stable" plant condition for NFPA 805 will not require boration of the RCS. Gravity insertion of the control rods into the reactor core will ensure reactivity control is achieved and maintained for Mode 3 (Hot Shutdown Condition).

4. Inventory makeup to the RCS may only be required to account for expected RCS leakage and minimal RCS shrinkage.

FCS has design features and procedures to ensure that an adequate source of borated inventory is provided for RCS inventory control in sustained Mode 3 (Hot Shutdown Condition) (i.e., RCS Page 21 FC~5~ NFPA RflR Trnn~itinn R,~nnrt inventory makeup from the SIRWT and/or the BASTs to maintain the RCS sub-cooled) utilizing the CVCS or the HPSI system.5. FCS has design features and procedures to ensure that an adequate source of heat input is maintained for RCS pressure control in sustained Mode 3 (Hot Shutdown Condition) (i.e., a minimum of 150kW of pressurizer heater input to maintain the RCS sub-cooled) utilizing available combinations of the backup pressurizer heaters (banks -1 and -4 are 225kW each, banks -2 and -3 are 150kW each). The backup pressurizer heaters are capable of being energized from emergency diesel generator power.The FCS NFPA 805 non-power operations assessment addresses fire risk during the"highest risk evolutions" which may occur in Mode 3 (Hot Shutdown Condition) (from the point at which the MCC breakers for the SDC suction valves, HCV-347 and HCV-348, are un-locked and closed, at which point spurious operation of these high/low pressure interface valves can occur due to fire damage to the valve control circuitry), Mode 4 (Cold Shutdown Condition), and/or Mode 5 (Refueling Shutdown Condition).

The"highest risk evolutions" considered in the FCS non-power operations assessment for these non-power operational modes are consistent with the information and technical approach provided in FAQ 07-0040.Precedent for a 24-hour "safe and stable" coping time as defined herein by FCS for NFPA 805 is established in LIC-93-0278, "NRC Generic Letter 88-20 Submittal for Fort Calhoun Station 'Individual Plant Examination for Severe Accident Vulnerabilities' (TAC No. .74412)", and NRC-96-0216, "Fort Calhoun Station, Unit No. 1 -Review of Individual Plant Examination (IPE) Submittal

-Internal Events (TAC-No. M74412)." These licensing documents discuss a 24-hour mission time(s) with respect to the FCS response to various postulated severe accidents.

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 C.2.21 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02 suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee's methodology.

The discussion below provides the methodology used to transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML1 03090602) 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.

Page 22 Om-qh-q P"hlir Pnimorni-4drirt P,,h~~~~~ir~~

R~~ lcti~ fF NFZPA RnlFi Trnnqition Rennv* Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth)." Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path.* Step 4: Evaluate the feasibility of the recovery actions." Step 5: Evaluate the reliability of the recovery actions.Results The review results are documented in FCS engineering analysis EA10-041, Recovery Action Feasibility Assessment, and FCS calculation FC07883, Fire Risk Assessment of FCS Variances from Deterministic Requirements of NFPA 805. Time critical operator actions required to mitigate a design basis event are included in the FCSG-56, Time Critical Action (TCA) program. PRA events are listed in the TCA program for tracking purposes, when human failure events are risk significant using the Maintenance Rule criteria of Risk Achievement Worth (RAW) of greater than 2.0 or a Risk Reduction Worth of greater than 1.005, and the time available/time to completion ratio is less than 2.0. 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 FCS is summarized below.As part of the NFPA 805 transition project, a review and evaluation of FCS susceptibility to fire-induced MSOs was performed.

The process was conducted considering guidance from NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038 Revision 3 (ML110140242).

The PWR Generic MSO list, WCAP-16933-NP Revision 0, dated April 2009 was utilized.The approach outlined in Figure 4-3 (based on 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 the NFPA 805 NSCA to include the MSOs of concern.Page 23 Omaha Public Power District FCS NFPA 805 Transition Reno* 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).Page 24 Omaha Public Power District FCS NFPA 805 Transition Renort Identify Potential MSOs of Concern* SSA Step 1

• 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 4 Pursue other resolution options Step 5 Document Results Figure 4-3 -Multiple Spurious Operations

-Transition Resolution Process (Based on FAQ 07-0038)Results Refer to Attachment F for the process used by FCS and the results from the process.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 includes the following determinations:

.The EEEE is not based solely on quantitative risk evaluations, Page 25 Omaha Public Power District FCS NFPA 805 Transition Report* 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 and NEI 04-02, 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 should be referenced where required and a brief description of the evaluated condition should be provided.* If requesting specific NRC approval for "adequate for the hazard" EEEEs, then EEEE should be referenced where required to demonstrate compliance and a detailed summary, including sufficient detail to allow the NRC staff to evaluate the EEEE should be provided.

At a minimum, the level of detail is expected to include: (1) a summary of each condition, (2) a summary of the evaluation of each condition, and (3) a summary of the resolution of each condition.

In all cases, the reliance on EEEEs to demonstrate compliance with NFPA 805 requirements should be documented in the LAR.Results The review results for EEEEs are documented in FCS Engineering Analysis EA10-063, Engineering Evaluation Review.In accordance with the guidance in RG 1.205 and NEI 04-02, EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in Attachments A and C as appropriate.

In addition, none of the transitioning EEEEs require NRC approval.4.2.3 Licensing Action Transition Overview of Evaluation Process The existing licensing actions (exemption requests / deviations

/ safety evaluations) review was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following:

• Determination of the bases for acceptability of the licensing action.* Determination that these bases for acceptability are still valid and required for NFPA 805.Page 26 Omaha Public Power District FCS NFPA 805 Transition Report Results Attachment K contains the results of the Licensing Action Review. Licensing actions identified as required post-transition will be transitioned into the NFPA 805 fire protection program. These licensing actions are considered compliant under 10 CFR 50.48(c).

OPPD is requesting that the licensing actions below (exemptions and their bases) be rescinded and transitioned to the new licensing basis under 10 CFR 50.48(a)and 50.48(c) as previously approved (NFPA 805, Section 2.2.7) and compliant with the new regulation.

Regardless of whether the compliance strategy for a given fire area is deterministic or performance-based, every exemption identified is being transitioned and credited for compliance under the NFPA 805 licensing basis. The following exemptions exist in plant licensing documentation and are reviewed in Attachment K: " Fire area 30: containment, lack of 20-foot separation free of intervening combustibles (July 3, 1985 and July 1, 1986)" Fire area 31: intake structure and pull boxes, lack of a one-hour fire barrier, lack of area-wide suppression, and lack of detection in pull box area (July 3, 1985)" Fire area 32: air compressor room, lack of a one-hour fire barrier (July 3, 1985 and July 1, 1986)" Fire area 34A: electrical penetration area, lack of area-wide suppression (July 3, 1985 and March 17, 1993)" Fire area 34B-1: electrical penetration area, lack of area-wide suppression (July 3, 1985)" Fire areas 36A, 36B, 36C: switchgear room, lack of three-hour rated barrier between redundant shutdown divisions (July 3, 1985)" Fire area 42: control room, lack of area-wide suppression in alternate shutdown area (July 3, 1985)" Fire area 30: RCP lube oil collection system, lube oil holdup tank capacity (December 20, 1988)" Fire area 30: RCP lube oil collection system, unprotected oil leakage sites (May 21, 1998)" Fire area 47: provision of repair procedures and materials for cold shutdown capability for redundant cold shutdown components within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (February 6, 2009)For the appropriate licensing actions (fire protection program exemptions), these requests are also included in Attachment 0, Orders and Exemptions.

In some cases prior NRC approval included in a licensing action may be unclear.OPPD requests that the NRC concur with their finding of prior approval for the following: " Separation of steam generator level and pressure, RCS temperature, and source range neutron flux monitoring instrumentation inside containment" Separation of RCS loop charging valves inside containment Page 27 Omaha Public Power District FCS NFPA 805 Transition ReDort" Separation of pressurizer auxiliary spray valves inside containment" Separation of pressurizer level instrumentation inside containment" Separation of pressurizer pressure instrumentation inside containment

• Separation of backup pressurizer heater groups inside containment

• Separation of raw water pumps, discharge valves, and strainers inside the intake structure" Separation of redundant trains inside the compressor room* Protection of redundant 480V MCCs inside the lower electrical penetration area* Protection of cable tray sections inside the switchgear rooms" Separation of raw water pumps, discharge valves, and strainers in pull boxes, underground cable duct bank and manhole vaults* Lack of full suppression in the control room The discussion of the prior approval, including appropriate reference documents, is provided in Attachment T.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 -Assemble documentation.

Gather industry and plant-specific fire area analyses and licensing basis documents.

.Step 2 -Document fulfillment of nuclear safety performance criteria." Assess accomplishment of nuclear safety performance goals. Document the method of accomplishment, in summary level form, for the fire area.* Document evaluation of effects of fire suppression activities.

Document the evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria." Perform licensing action reviews. Perform a review of the licensing aspects of the selected fire area and document the results of the review. See Section 4.2.3.* Perform existing engineering equivalency evaluation reviews. Perform a review of existing engineering equivalency evaluations (or create new evaluations) documenting the basis for acceptability.

See Section 4.2.2.* Pre-transition OMA reviews. Perform 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.Page 28 Omaha Public Power District FCS NFPA R05 Transitinn Ppnnrt Step 3 -VFDR Identification, characterization, and resolution considerations.

Identify variances from the deterministic requirements of NFPA 805, Section 4.2.3. Document variances as either a separation issue or a degraded fire protection system or feature.Develop VFDR problem statements to support resolution.

Step 4 -Performance-Based evaluations (Fire Modeling or Fire Risk Evaluations)

See Section 4.5.3 for additional information.

Step 5 -Final Disposition.

• Document final disposition of the VFDRs in Attachment C (NEI 04-02 Table B-3)." For recovery action compliance strategies, ensure the manual action feasibility analysis of the required recovery actions is completed.

Note: if a recovery action cannot meet the feasibility requirements established per NEI 04-02, then alternate means of compliance must be considered." Document the post transition NFPA 805 Chapter 4 compliance basis.Step 6 -Document required fire protection systems and features.

Review 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 protection systems and features are subject to the applicable requirements of NFPA 805 Chapter 3.Page 29 Omaha Public Power District FCS NFPA 805 Transition Report Identify INITIAL Variances From Deterministic Requirements of NFPA 805 § 4.2.3 (B-3 Table)Document Final Disposition of VFDR Compliance options include: Accept As Is Require FP systems/features Require Recovery Action Require Programmatic Enhancements Require Plant Modifications (B-3 Table)Figure 4-4 -Summary of Fire Area Review[Based on FAQ 07-0054 Revision 1]Page 30 Omaha Public Power District FCS NFPA 805 Transition Renort 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.* Licensing Actions -Specific references to exemption requests/safety evaluations that will remain part of the post-transition licensing basis. A brief description of the condition and the basis for acceptability of the licensing action should be provided.

Attachment T contains items for which FCS is requesting concurrence of prior approval.* EEEE -Specific references to EEEEs 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 4.5.3 for a discussion of the performance-based approach.Risk criteria information provided in the Required Fire Protection Systems and Features tables in Attachment C originates from FC07823. All other information provided in Attachment C originates from EA10-044 (NEI 04-02 Table B-3).4.3 Non-Power Operational Modes 4.3.1 Overview of Evaluation Process FCS implemented the process outlined in NEI 04-02 and FAQ 07-0040, Non-Power Operations Clarifications.

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 involves the following steps:* Review the existing Outage Management Processes.

• Identify Equipment/Cables:

• Review plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and Page 31 Omaha Public Power District FCS NFPA 805 Transition Report Identify cables required for the selected components and determine their routing." Perform Fire Area Assessments (identify pinch points -plant locations where a single fire may damage all success paths of a KSF)." Manage pinch-points associated with fire-induced vulnerabilities during the outage.The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2.Page 32 Omaha Public Power District FCS NFPA 805 Transition Reoort Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points Page 33 Omaha Public Power District FCS NFPA 805 Transition Renort NO 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 Fire Protection NO Defense-in-Depth Actions Fire Protection NO Defense-in-Depth Actions Implement Contingency Plan for Specific KSF Figure 4-6 Manage Pinch Points Page 34 Omaha Public Power District FCS NFPA 805 Transition Rignort 4.3.2 Results of the Evaluation Process FCS outage management processes were reviewed.

It was determined that an explicit definition of higher risk evolutions consistent with the guidance of FAQ 07-0040 did not exist in the FCS outage management processes.

In lieu of an explicitly defined set of plant conditions as a higher risk evolution (HRE) period, an evaluation was performed of the evolutions that FCS performs during an outage. This evaluation determined that the evolutions performed, and the plant conditions experienced by FCS during an outage are consistent with the Plant Operational States (POS) discussed in FAQ 07-0040.FCS engineering analysis EA10-042 discusses the POS reviewed, and the plant systems and equipment selected to provide the key safety functions (KSF) of Inventory Control, Decay Heat Removal Capability, Reactivity Control, Containment Closure, and support functions (Process Cooling, and Electrical Power).The selected equipment was logically associated with the supported KSF(s). Power supplies, interlocks, and supporting equipment were logically associated with their parent component, as needed. These data relationships are stored electronically for use with an analytical software tool.For components which had not been previously cable selected for other NFPA 805 tasks, cable selection was performed per the Nuclear Safety Methodology.

The cables necessary to support the selected function of a component were selected and analyzed for fire impact.In accordance with FAQ 07-0040, any area experiencing fire damage which eliminates all success paths for a KSF (without recovery actions outside the main control room) is considered a 'pinch point'.FCS engineering analysis EA10-042 contains the NPO fire area assessment, the identified

'pinch points', and the credited actions consistent with FAQ 07-0040 to reduce fire risk. Fire modeling was not used to eliminate any fire area from being a pinch point.As part of the NPO fire area assessments, FCS has credited the completion of plant modifications from the at-power NFPA 805 NSPC analysis (e.g., additional fusing installed on electrical circuitry, etc.) that will be completed as part of LAR implementation (see Attachment S).The list of credited actions specified in FCS engineering analysis EA10-042 considers the following actions from FAQ 07-0040:* Prohibition or limitation of hot work in fire areas during periods of increased vulnerability." Verification of operable detection and/or suppression in the vulnerable areas." Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability." Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position).

Page 35 Omaha Public Power District FCS NFPA 805 Transition Reoort* Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability.

• Use of recovery actions to mitigate potential losses of KSFs.* Identification and monitoring in-situ ignition sources for "fire precursors" (e.g., equipment temperatures).

• Reschedule the work to a period with lower risk or higher DID.Attachment D, Table D-1, provides additional detail for the FCS NPO transition.

Implementation of the NPO fire area assessment results into the FCS outage management processes will be completed as part of LAR implementation. (See Attachment S).4.4 Radioactive Release Performance Criteria 4.4.1 Overview of Evaluation Process The review of the Fire Protection Program against NFPA 805 requirements for fire suppression related radioactive release was performed using the methodology contained in Nuclear Energy Institute (NEI) 04-02 and subsequent guidance provided in NFPA 805 Task Force FAQ 09-0056. The methodology consisted of the following: " "Screen-in" fire zones based on the potential to contain contaminated materials during all plant operating modes, including full power and non-power conditions.

The screening process considers input from Radiation Protection (RP) personnel and review of FCS pre-fire plans. The evaluation focuses on radioactive release to any unrestricted area due to firefighting activities only; radioactive release due to potential fuel cladding damage is not evaluated.

The nuclear safety goal, nuclear safety objectives, and nuclear safety performance criteria specified in NFPA 805 require the prevention of fuel cladding damage. As such, radiological release due to fuel damage does not require a separate examination since no such damage is assumed to occur without violating the basic requirements of NFPA 805." Review pre-fire plans and fire brigade training materials to identify FPP elements (e.g., systems / components

/ procedural control actions / flow paths, etc.) that are being credited to meet the radioactive release goals, objectives, and performance criteria during all plant operating modes, including full power and non-power conditions." Review engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents). This review shall include all plant operating modes (including full power and non-power conditions).

Otherwise, provide a bounding analysis, quantitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of radioactive effluents specified in the unit's Offsite Dose Calculation Manual (ODCM) are met.Page 36 nmnhn P"hlit- Pnimnr r)ic.,trh-f P.qAiPPA RflF Trnn'zitin Pion,, FCS engineering analysis EA10-043 details the methodology for the screening process and review of pre-fire plans, fire brigade training materials, and engineering controls.4.4.2 Results of the Evaluation Process The radioactive release review determined the FP program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the implementation items identified in Attachment S.The site specific review of the direct effects of fire suppression activities on radioactive release is summarized in Attachment E.4.5 FPRA 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 (FPRA), 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 FPRA Development and Assessment In accordance with RG 1.205, the FCS FPRA model was developed in accordance with the requirements of Part 4 "Internal Fires at Power Probabilistic Risk Assessment Requirements," of the combined ASME/ANS 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 the FPRA Standard).

The model was developed using guidance in NUREG/CR-6850 and its supplemental documents.

The FCS FPRA was peer reviewed by independent team industry experts as required by RG 1.200 prior to a risk-informed submittal.

The resulting FPRA model is used to support FREs during the transition process.Section 4.5.1.1 describes the FCS internal events PRA model. Section 4.5.1.2 describes the FCS FPRA model. Section 4.5.1.3 describes the results of the FPRA peer review and resolution of its findings, and Section 4.5.1.4 describes insights gained from the FPRA.4.5.1.1 Internal Events PRA Quality Revision 11 of the FCS IE PRA was the starting point for the FCS FPRA. The FCS IE PRA has undergone several peer reviews and self-assessments since 1999. The reviews have confirmed that the FCS internal events PRA complies with Regulatory Guide 1.200, Revision 2 and meets Capability Category II for most SRs of ASME/ANS RA-Sa-2009 Part 2 and Part 3. Attachment U provides additional information on the internal events model quality.4.5.1.2 FPRA Quality The FCS FPRA was developed in accordance with the requirements of ASME/ANS RA-Sa-2009 Part 4 and using the guidance in NUREG/CR-6850.

Although NUREG/CR-Page 37 Omaha Public Power District FCS NFPA 805 Transition ReDort 6850 was the primary guidance document, the FCS FPRA selectively applied guidance, methods, and data from the following documents: " EPRI TR-100370, Fire-Induced Vulnerability Evaluation (FIVE), Electric Power Research Institute (EPRI), Palo Alto, CA, May 1992* EPRI TR-1 00443, Methods of Quantitative Fire Hazard Analysis, EPRI, Palo Alto, CA, May 1992* EPRI 1003111, Fire Events Database and Generic Ignition Frequency Model for U.S. Nuclear Power Plants, EPRI, Palo Alto, CA, 2001." EPRI 1016735, Fire PRA Methods Enhancements:

Additions, Clarifications, and Refinements to EPRI 1019189, EPRI, Palo Alto, CA, December 2008." EPRI 1019259, Fire Probabilistic Risk Assessment Methods Enhancements Supplement 1 to NUREG/CR-6850 and EPRI 1011989, EPRI and United States Nuclear Regulatory Commission, September 2010." Macleod, D., Kolonauski, L., and Jan Grobbelaar, Simplified HRA Process for Internal Fire Analysis, ANS PSA 2008 Topical Meeting, Knoxville, Tennessee, September 7-11, 2008.* NUREG-1805, Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program. December, 2004. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, DC." NUREG-1921 (EPRI TR-1019196), Draft Report for Public Comment, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, EPRI, Palo Alto, CA, and USNRC, Office of Nuclear Regulatory Research, Washington, D.C.: November 2009.In addition, the FCS FPRA applies four methods outside of NUREG/CR-6850 that have not been formally "approved" by the NRC Frequently Asked Questions (FAQ) process or the peer review process EPRI has set up for new FPRA methods. The first method involves a generic severity factor / non-suppression probability for electrical cabinets, and this method was submitted to the EPRI New Methods Peer Review Panel in April 2011. The second method involves pump fire frequency apportioning that assigns a higher frequency to normally operating pumps and lower frequency to standby pumps.The third method involves a diesel generator generic severity factor / non-suppression probability accounting for most diesel generator fire events occurring during surveillance testing and the prompt suppression by test personnel.

These second and third methods are included in the industry FPRA refinement efforts documented in MLl 10210990,"Roadmap for Attaining Realism in Fire PRAs." Finally, the fourth method involves application of a draft version of FAQ 08-0050, "Non-Suppression Probability," dated May 30, 2008. This draft version of the FAQ was implemented because it was the most current version when the relevant OPPD task was initiated.

The more recent version of FAQ 08-0050 is documented in NUREG/CR-6850, Supplement 1, issued September 2010.Page 38 NFPA Rfl.'~ Trnn~itirn R,~ncwt OPPD has performed sensitivity studies that demonstrate FCS satisfies the NFPA 805 acceptance criteria, even if the four alternate methods identified above are replaced by methods within the guidance of NUREG/CR-6850 and its Supplement

1. Although the acceptance criteria would still be met, the alternate methods are retained in the model to help focus on, and not skew, risk significant insights and to better align with similar methods currently under development and review by the industry and NRC. Refer to Section W.2 for results of the sensitivity studies.Fire Model Utilization in the Application Fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2).

RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis.The following fire models were used to develop the FPRA: Algebraic equations implemented by the NRC Fire Dynamics Tools (Reference NUREG-1805) and the EPRI Fire-Induced Vulnerability Evaluation methodology were used to characterize flame radiation, flame height, plume, ceiling jet, and hot gas layer for a variety of ignition source types and heat release rates. These calculations were primarily performed to identify zones of influence, calculate severity factors, and determine times to target damage in support of non-suppression probability calculation." The NIST Fire Dynamics Simulator was used to assess main control room habitability during fire events.Attachment J discusses the acceptability of these fire models.4.5.1.3 Results of FPRA Peer Review The FCS FPRA has received a formal industry peer review against the Part 4 requirements of ASME/ANS RA-Sa-2009 and in accordance with the peer review guidelines of NEI 07-12. This peer review was conducted September 27 -October 1, 2010 by a diverse group of industry experts, collectively representing all skill sets required to critically review a FPRA. The peer review covered all aspects of the FCS FPRA model and the administrative processes used to maintain and update the model.The review generated specific recommendations for model, documentation, and process improvements, and these recommendations are documented in the form of Facts and Observations (F&Os) in the peer review report.The FCS peer review report documents that for 160 of 182 total supporting requirements (SRs), the FCS FPRA model was assessed as either 'Met', CC-Il, CC-Il/111, CC-Ill, or 'Not Applicable'.

22 of 182 total SRs were assessed as 'Not Met' or 'CC-I'. Note that the acronym "CC" refers to the "Capability Categories" associated with each SR. The FPRA model and documentation have been revised such that all but one (FSS-D7) of these 22 SRs can now be classified as either 'Met', CC-Il, CC-Il/Ill, or CC-II1. Regarding FSS-D7, OPPD's position is that this SR is met at CC-I, and this is Page 39 Omaha Public Power District FCS NFPA 805 Transition Report acceptable for the NFPA 805 application as discussed in Table V-3. See Attachment V for additional information on OPPD FPRA quality.While the FCS FPRA met 18 SRs at the CC-Ill or CC-Il/Ill levels per letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. L. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 7.17 PRA Peer Review History," dated April 1, 2011 (CFTC-11-95), all supporting requirements were reviewed against the VFDRs documented in EA10-044, and no specific SRs were identified as needing to meet CC-Ill to appropriately assess their risk-significance.

An example of when an SR might need to meet CC-Ill would be to assess a VFDR that involves simultaneous spurious actuation of more components than required to meet CC-Il of the relevant Equipment Selection (ES) SRs.Attachment V contains a summary of the FPRA peer review F&Os and their resolution by OPPD.4.5.1.4 Risk Insights Risk insights were revealed and documented as part of FCS FPRA development.

Attachment W documents a summary of the FCS FPRA insights as well as the fire initiating events that collectively represent 95% of the calculated FCS fire risk.4.5.2 Performance-Based Approaches NFPA 805 Section 4.2.4 identifies the following two approaches for implementing performance-based analyses:* Fire Modeling (NFPA 805 Section 4.2.4.1)." FRE (NFPA 805 Section 4.2.4.2).The FCS NFPA 805 transition implemented the FRE approach per NFPA 805 Section 4.2.4.2 to evaluate the risk significance and acceptability of VFDRs. The fire modeling approach, per NFPA 805 Section 4.2.4.1, was not utilized for the transition.

4.5.2.1 Fire Modeling Approach The fire modeling approach, per NFPA 805 Section 4.2.4.1, was not utilized for the FCS NFPA 805 transition.

4.5.2.2 Fire Risk Approach Overview of Evaluation Process The FCS NPFA 805 transition performed Fire Risk Evaluations (FREs) to assess the risk significance and acceptability of VFDRs of NFPA 805 Section 4.2.3. These FREs were performed using the industry guidance summarized in Table 4-1.Page 40 Omaha Public Power District FCS NFPA 805 Transition ReDort Table 4-1 FRE 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 FRE (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 FRE 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)The general FRE process is summarized in the following steps and depicted in Figure 4-7: Step I -Prepare for the FRE" Define the VFDRs -The definition of each VFDR should include a problem statement, the section of NFPA 805 that is not met, the type of VFDR (e.g., separation issue or degraded fire protection system), and the proposed evaluation per applicable NFPA 805 section.* Prepare for the Evaluation

-Each VFDR is first reviewed by a team of stakeholders and experts to understand the scope of each VFDR, understand the FPRA modeling treatment of each VFDR, identify and resolve any discrepancies, and develop an overall FRE strategy for each VFDR. Depending on the scope and complexity of each VFDR, the team may include the post-fire safe shutdown engineer, fire protection engineer, FPRA engineer, and supplemental support as necessary.

Step 2- Perform the FRE The total ACDF and ALERF for each fire area containing VFDRs is calculated.

For a given fire area, the reported ACDF and ALERF represent the total change in risk as a result of the contribution of all VFDRs within that fire area. The FPRA engineer coordinates this effort with the post-fire safe shutdown engineer, fire protection engineer, and supplemental support as necessary.

Recovery actions credited by the post-fire safe shutdown analysis are also reviewed for their potential to adversely affect risk, and any adverse contribution identified is either eliminated (i.e., by revising or eliminating the recovery action) or incorporated into the fire risk calculations.

Page 41 Omaha Public Power District FCS NFPA 805 Transition Report Step 3 -Review the Acceptance Criteria* The resulting ACDF and ALERF for each VFDR fire area are compared to the quantitative acceptance criteria in RG 1.174, as clarified by RG 1.205 Regulatory Position C.2.2.4.* If the VFDR satisfies the quantitative acceptance criteria, a review is performed to assess whether adequate defense-in-depth and safety margin are maintained, per the guidance in NEI 04-02. This review may result in additional stipulations (e.g., crediting recovery actions, plant modifications, procedure modifications) for the VFDR to be considered acceptable.

Page 42 Omaha Public Power District FCS NFPA 805 Transition ReDort Omaha Public Power District FCS NEPA 805 Transition Renort Prepare for Fire Risk Evaluation Perform Fire Risk Evaluation Review of Acceptance Criteria Figure 4-7 -FRE Process for NFPA 805 Transition (Based on FAQ 07-0054 Revision 1)Page 43 Omaha Public Power District FCS NFPA 805 Transition ReDort Results of Evaluation Process Disposition of VFDRs The FCS NFPA 805 transition has identified a number of VFDRs of NFPA 805 Section 4.2.3. Each of these VFDRs was dispositioned using the FRE process.Each VFDR dispositioned using the FRE process was assessed against the ACDF and ALERF acceptance criteria, maintenance of defense-in-depth, and maintenance of safety margin per NEI 04-02 and RG 1.205. Attachment C identifies each FCS VFDR and summarizes the associated FRE conclusions.

Attachment W summarizes the ACDF and ALERF for each VFDR fire area., As a result of the FREs, several recovery actions were credited for defense-in-depth.

As summarized in Table W-5, these actions were qualitatively reviewed and determined to be risk beneficial and not pose any net adverse effect on plant risk.Following completion of transition activities, including planned plant modifications and programmatic changes, FCS will be compliant with 10 CFR 50.48(c).Risk Change Due to NFPA 805 Transition RG 1.205 Section C.2.2.4 requires that net change in risk, for the overall plant, associated with implementing NFPA 805 be reported.

Attachment W provides this net ACDF and ALERF associated with the NFPA 805 transition, and they are within the Region II acceptance criteria of RG 1.174.4.6 Monitoring Program NFPA 805 Section 3.2.3(3) requires that procedures be established for reviews of the fire protection program related performance and trends. NFPA 805, Section 2.6 requires a monitoring program that in part is to establish acceptable performance levels and a method to monitor and assess the performance of the fire protection program.The NFPA 805 requirements for reviews of programs related to performance and trending is provided under the NFPA 805 monitoring program.The monitoring program will be implemented after the safety evaluation issuance as part of the FP program transition to NFPA 805. In order to assess the impact of the transition to NFPA 805 on the current monitoring program, the OPPD FP program documentation such as the maintenance program processes, FP program implementing procedures, and plant change processes will be reviewed.

Sections 4.5.3 and 5.2 of the NEI 04-02 will be used during the review process and that process is described in the following sections.The following scope will be documented appropriately:

• The scope of SSCs and programmatic elements to monitor." The levels of availability, reliability, or other criteria for those elements that require monitoring.

Development and implementation of the NFPA 805 monitoring program for FCS will be completed as part of LAR implementation. (See Attachment S).Page 44 Omaha Public Power District FCS NFPA 805 Transition Report 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states: "A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria.

Monitoring shall ensure that the assumptions in the engineering analysis remain valid." The intent of the monitoring review is to confirm (or modify as necessary) the adequacy of the existing surveillance, testing, maintenance, compensatory measures, and oversight processes for transition to NFPA 805. This review will consider the following:

The adequacy of the scope of systems and equipment within existing plant programs, i.e., the necessary FP systems and features and nuclear safety capability equipment (NFPA 805 Section 1.5.1) are included.The performance criteria for the availability and reliability of FP systems and features relied on to demonstrate compliance.

The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence.

4.6.2 Overview

of Post-Transition NFPA 805 Monitoring Program This section describes the overall Post-Transition NFPA 805 Monitoring Program process. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring process will be conducted in four phases." Phase 1 will determine the scope which includes fire protection systems &features and nuclear safety capability equipment." Phase 2 will establish performance criteria." Phase 3 will determine risk significant fire protection program and defense-in-depth elements using criteria established in Phase 2." Phase 4 will implement the program after the scope and criteria are established.

Performance and availability monitoring criteria established in Phase 2 will be applied to the risk significant fire protection systems and features and a tracking program will be used on the remaining NFPA 805 required fire protection systems and features.

This process will result in development of a program that reviews the fire protection program performance and identifies trends in performance.

The reviews will be based on specific performance goals established to measure the effectiveness of the fire protection program. Monitoring will ensure that assumptions in engineering analysis remain valid. The monitoring program will be documented in an administrative process (i.e., program manual or directive) that provides the process and sets clear guidelines to consistently measure the performance of the fire protection program.The phases of the monitoring process are described as follows: Page 45 Omaha Publicr Power flisqfrict SC NFPA 80l5 Transition Re.gnor Phase 1 -Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements shall be included in the NFPA 805 monitoring program: 0 Fire Protection Structures, Systems, and Components" Fire protection systems and features required by the NSCA" Fire protection systems and features modeled in the Fire PRA" Fire protection systems and features required by Chapter 3 of NFPA 805* Fire Protection Programmatic Elements* Key Assumptions in Engineering Analyses As a minimum these fire protection features and systems will be included in the existing fire protection surveillance and system/program health programs.

The following process is suggested to determine those fire protection systems and features that may require additional monitoring beyond normal surveillance activities.

1. Fire Protection Structures, Systems, and Components Monitoring of SSCs that are required to demonstrate compliance with NFPA 805 is required.

These SSCs may include Detection and Alarm Systems, Fire Suppression Systems, Water Supply, Hydrants, and Valves, Fire Pumps, Stand Pipes, Hose Stations, and Hoses, or Fire Barriers, among others. Only those fire protection systems and features required by the NSCA or modeled in the Fire PRA would be considered in scope for the additional monitoring of the NFPA 805 program.2. Monitoring of 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." Programmatic aspects include: " Transient Combustible Control; Transient Exclusion Zones" Hot Work Control; Administrative Controls" Fire Watch Programs; Program compliance and effectiveness

• Fire Brigade; Response Times Fire Protection Health Reports, Self-Assessments, regulator and insurance company reports provide inputs to the monitoring program. The monitoring of programmatic elements and program effectiveness may be performed as part of the management of engineering programs.

This monitoring is more qualitative in nature since the programs do not lend themselves to the numerical methods of reliability and availability.

These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements.

Page 46 nmnhnP"h1if, Pnwor Mctrirf X: _4Z AJIX:IA %P1l; Trmnoifirnn P~n nr I VLI IA I t~kIILI LVVLS ~IILIIt/.

I ~ I'II ri t/t I V.1 '~lLISI I S..JJLIV V 3. Monitoring of Key Assumptions in Engineering Analyses The assumptions of the Fire PRA are the primary drivers of the need for monitoring levels of reliability and availability of the SSCs utilized in the risk informed performance based program. These SSC's are generally broken down into two groups, the NSCA (and PRA Internal Events) SSCs and the fire protection systems and features SSCs. Other analytical assumptions from the NSCA, Non-Power Operations and Radioactive Release evaluations may also increase the scope of Fire Protection SSCs or programmatic elements to be reviewed.

The NFPA 805 Monitoring program shall be used to monitor the performance of these Fire Protection SSCs at either the component or the functional level.NSCA and PRA internal events equipment and systems are generally monitored by the Maintenance Rule. It is anticipated that in most cases, for the NSCA type components, the existing Maintenance Rule performance goals will be bounding.Any NSCA equipment and systems not considered under the Maintenance Rule should be reviewed for inclusion in the Maintenance Rule.Phase 2 -Establishing Risk Criteria Phase 2 of the process is establishing the risk significant criteria and screening for the FP SSCs and programmatic elements within the NFPA 805 monitoring scope. This may be accomplished at the component, programmatic element, and/or functional level.Since risk is evaluated at the compartment level, criteria must be developed to determine those compartments (or analysis units) for which the FP SSCs are considered risk significant.

The Fire PRA is the primary tool used to establish risk significance criteria and performance bounding guidelines.

The screening thresholds used to determine risk significant fire compartments are those that meet the following example criteria: " Risk Achievement Worth (RAW) of the monitored parameter

> 2.0 (AND) either* Core Damage Frequency (CDF) x (RAW) > 1.OE-7 per year (OR)* Large Early Release Frequency (LERF) x (RAW) > 1.OE-8 per year High Safety Significant Fire Protection SSCs are those that meet or exceed the risk significant fire compartment screening criteria, and all required FP SSCs, programmatic elements and /or functions are included for each fire compartment.

Low Safety Significant Fire Protection SSCs are those that do not meet the risk significant fire compartment screening criteria and are monitored via existing programs/processes.

Additionally, the Expert Panel or reviewer may include other fire compartments (and required FP SSCs, programmatic elements and /or functions) that are not risk significant (per the Fire PRA screening criteria) but are included based on plant specific history and/or operational considerations.

Page 47 Omaha Public Power District FCS NFPA 805 Transition Renort As an alternative to including the required FP SSCs, programmatic elements and /or functions in the entire fire compartment, fire protection equipment and features can be included based on a smaller analysis unit (ignition source). Basis needs to be provided when using this approach to ensure adequate monitoring is provided.EXAMPLE (from NFPA 805 FAQ): For a plant, the power block definition included the turbine building.

The Fire PRA had made the entire turbine building (four floors, open to the outside, approximately 52,800 square feet) one fire compartment.

Values for CDF and LERF are greater than the threshold, so this fire compartment is screened into the monitoring program. There are four significant fire sources identified (for CDF and LERF) for this fire compartment.

Two fire sources are located in the general service switchgear room on the south side of the 261' elevation, one fire source is located on the northeast corner on the 261' elevation, and one fire source is in the electrical room on the south side of the 286' elevation.

These four fire sources would contribute 350 detectors, 18 detector channels, 16 sprinkler valves, and ten manual pull stations into the scope of systems requiring additional monitoring.

When just the impact from the four sources within the compartment is considered, the monitored equipment is 42 detectors, three detector channels, one sprinkler valve, and one manual pull station. This accounts for an almost 90% reduction in quantity of monitored equipment while still focusing on the important fire scenarios.

The more practical and realistic approach to this particular fire compartment would be to evaluate each of the four significant fire sources, determine exactly what equipment would mitigate the impact of the four significant fire sources, and to only include that equipment in the monitoring program.Phase 3 -Risk Determination Phase 3 consists of utilizing the Fire PRA, or other processes as appropriate, to determine target values of reliability and availability for the High Safety Significant FP SSCs, programmatic elements and /or functions using the criteria established in Phase 2.Failure Criteria are established by the Expert Panel or evaluation based on the required FP SSCs, programmatic elements and/or functions assumed level of performance in the supporting analyses.

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 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. Documentation of the Monitoring Program failure criteria and action level targets will be contained in the Expert Panel Meeting Minutes or other documented evaluation.

Phase 4 -Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established.

The corrective action process will be used to address performance of Fire Protection SSCs that do not meet Performance Criteria.Page 48 Omaha Public Power District FCS NFPA 805 Transition ReDort For High Safety Significant Fire Protection SSCs that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, the Expert Panel approves the Corrective Action criteria and action level adjustment if more than usual monitoring is warranted.

A periodic assessment should be performed (e.g., at a frequency of approximately every two to three operating cycles), taking into account, where practical, industry wide operating experience.

This may be conducted as part of other established assessment activities.

Issues that should be addressed include: " Review Systems with Performance Criteria.

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

Page 49 Omaha Public Power District FCS NFPA 805 Transition Reoort Methdolgy Phaspng Scoping Phase 2 Screening Using Risk Criteria Phase 3 Risk Target Va Determination ue Phase 4 Monitoring Implementation NFPA 805 Monitoring Process (FAQ 10-0059)Page 50 Omaha Public Power District FCS NFPA 805 Transition Rei)ort 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, FCS has documented analyses to support compliance with 10 CFR 50.48(c).

The analyses are being performed in accordance with OPPD's processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail are 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 FCS "Fire Protection Design Basis Document" concept described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 will be created / revised as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation.

Appropriate cross references will be established to supporting documents as required by OPPD processes.

The development of the FP basis document will be completed as part of LAR implementation. (See Attachment S). Figure 4-8 depicts the planned post-transition documentation and relationships.

Page 51 Omaha Public Power District FCS NFPA 805 Transition Report.. .*. .............

Z o uL -h s.......NFPA 805 DOCUMENTS NSCA DATABASE I i lM-Power PRA Equipment Equipment and and Data Dt NSCA ANALYSIS FSO and OMA Treatm/enuts Melo/sut NSCA SUPPORTING INFO Manual Action Feasibility Assessment

[ B2TbJ t [ B-Tata Calculations I SuplnOort haCA HHIF reatment Non-Power Mode NSCA Treatment Non-Po ainAna I NFPA 805 FIRE RISK EVALUATIONS Fire Risk FHA DATA Comusible F e s~I IFeatures Data Calculations J .............

I....................................

FHA SUPPORT DOCUMENTATION FP ysems1 EngI I nering FPq nms andcy Feature DB~s Evaluations Release Review Analysis SRevisel LicenseJ Condition IFSA Revised UFA FIRE SAFETY ANALYSIS (DOD)On a Fire Area Basis-Fire Area Description

• FNA Database Information

-Nuclear Safety Performance Criteria Compliance Summary (NE 04-02 B.3 Table Resulta)* Non-Power Evaluations Results Summary-Radioactive Release Summary On a Generic Basis-B-1 Table Results Radioactive Release Wonitoring M Monitoring Program FIRE PRA Plant Partitioning Component Seleclion Cable Selection Qualitative Screening Fire Risk Modol Ignition Frequencies Scoping Fire Modeling Detailed Circuit Analysis Circuit Failure Likelihood Analysis Detailed Fire Modeling Fire HRA Seismic-Fire Interactions Quantification Uncertainty Analysis Bold text Indicates new NFPA 805 documents"SSEL includes the entire population of equipment I cables (NSCA. NPO. PPA)Figure 4-8 -NFPA 805 Planned Post-Transition Documents and Relationships Page 52 Omaha Public Power District FCS NFPA 805 Transition ReL)ort 4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 of NFPA 805 Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) is subject to OPPD 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 FP 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), Change Evaluation A.2.4.4, D.5 NEI 04-02 5.3, Appendix B, Appendix I, Change Evaluation, Change Evaluation Appendix J Forms (App. 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 4-9:* Defining the Change" Performing the Preliminary Risk Screening" Performing the Risk Evaluation" Evaluating the Acceptance Criteria Change Definition The Change Evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the Licensing Basis (NFPA 805 Licensing Basis post-transition).

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

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

Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-Page 53 Omaha Public Power District FCS NFPA 805 Transition Report 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 the change is not screened out through the screening process above, then proceed with the Risk Evaluation step.Risk Evaluation The screening is followed by engineering evaluations that may include fire modeling and risk assessment techniques.

The results of these evaluations are then compared to the acceptance criteria.

Changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework.

The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition.

The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature.The risk evaluation involves the application of fire modeling analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below.Acceptability Determination The Change Evaluations are assessed for acceptability using the ACDF (change in core damage frequency) and ALERF (change in large early release frequency) criteria from the license condition.

The proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.

Page 54 Omaha Public Power District FCS NFPA 805 Transition ReDort i Is the Yes Preliminary Risk Screening (5.3.3)-----------------------------------

Risk Evaluation (5.3.4)Initial Evaluat'on Screens? Yes-00 (5.3.4.1 & 2)PRA Capability Category Assessment No Fire PRA Detailed Capability Evaluation Category Assessment (5.3.4.3)-----------------------------------Acceptance Criteria (5.3.5)No C 6CDF &A LERF Change CK?Yes DID B and No SM OK?Yes No Docurr 111ýFigure 4-9 Plant Change Evaluation

[NEI 04-02 Figure 5-1]Note references in Figure refer to NEI 04-02 Sections Page 55 Omaha Public Power District FCS NFPA 805 Transition Report The FCS Fire Protection Program configuration is defined by the program documentation.

To the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses, and Fire Protection Program License Basis Reviews will be utilized to maintain configuration control of the Fire Protection program documents.

The configuration control procedures which govern the various FCS documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.

Several NFPA 805 document types such as: NSCA Supporting Information, Non-Power Mode NSCA Treatment, etc., generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. The new procedures will be modeled after the existing processes for similar types of documents and databases.

System level design basis documents will be revised to reflect the NFPA 805 role that the system components now play. The development of new control procedures and processes for new documents and databases created as a result of the transition to NFPA 805 will be completed as part of LAR implementation. (See Attachment S). The update of system level design basis documents to reflect the NFPA 805 role that the system components now play will be completed as part of LAR implementation. (See Attachment S).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, and/or 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 Section 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 FCS 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.

Page 56 Omaha Public Power District FCS NFPA 805 Transition Rei.ort 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality FCS will maintain the existing Fire Protection Quality Assurance program.During the transition to 10 CFR 50.48(c), FCS performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.Fire PRA Quality Configuration control of the Fire PRA model will be maintained by integrating the Fire PRA model into the existing processes used to ensure configuration control of the internal events PRA model. This process, which is discussed in OPPD procedure PED-SEI-37, "Probabilistic Risk Assessment Configuration Control," complies with ASME/ANS RA-Sa-2009 Section 1-5 and ensures that OPPD maintains an as-built, as-operated PRA model of the plant. The fire PRA configuration control process has been peer reviewed in accordance with NEI 07-12 and identified as a 'Strength' in the peer review report. This configuration control 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.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.

OPPD specifically requires that the calculations and evaluations in support of the NFPA 805 LAR be performed within the scope of the QA program. Quality procedure PED-QP-3, "Calculation Preparation, Review and Approval," requires configuration control for engineering calculations.

Quality procedure PED-QP-5, "Engineering Analysis Preparation, Review and Approval," requires that Fire Protection and Safe Shutdown Analyses have a review and an independent review, but not an independent design verification.

FCS Fire PRA documents are subject to the configuration control process for the PRA model per PED-SEI-37.

Specific Requirements of NFPA 805 Section 2.7.3 NFPA 805 Section 2.7.3.1 -Review Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with OPPD procedures.

PED-QP-3 requires configuration control for engineering calculations.

PED-QP-5 requires that Fire Protection and Safe Shutdown Analyses have a review and an independent review, but not an independent design verification.

FCS Fire PRA documents are subject to the configuration control process for the PRA model per PED-SEI-37.

Page 57 Omaha Public Power District FCS NFPA 805 Transition Rei3ort 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) are used and were used 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.For personnel performing fire modeling or Fire PRA development and evaluation, OPPD develops and maintains qualification requirements for individuals assigned various tasks. Position specific training 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 development of these qualification requirements and position specific training will be completed as part of LAR implementation. (See Attachment S).NFPA 805 Section 2.7.3.5 -Uncertainty Analysis Uncertainty analyses were performed as required by Section 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.

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 Table 4-3. The table provides the following information from the NEI 04-02 Table B-3: Fire Area: fire area identifier.

==

Description:==

fire area description.

NFPA 805 Regulatory Basis: Post-transition NFPA 805 Chapter 4 compliance basis (Note: Compliance is determined on a fire area 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 (ERFBS), 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 Page 58 Omaha P"hlir Pnwgr Di-qtrirt P('.q AIPPA Rn.JS Trnn-citinn Aann flm~hi~ PhIir~ PAw,~r flLc~trk~t It2~ AI~PA Rfl+/-S Tr~.n~itinn PI~nArf engineering evaluations, licensing actions, or as part of the reviews of the NEI 04-02 Table B-I process. The information is provided on an area basis. The basis for the requirement of the fire protection system / feature is designated as follows:* S -Separation Criteria:

Systems / Features required for Chapter 4 Separation Criteria in Section 4.2.3* L -LA Criteria:

Systems / Features required for acceptability of NRC approved Licensing Actions (i.e., Exemptions

/ Deviations

/ Safety Evaluations) (Section 2.2.7)* E -EEEE Criteria:

Systems / Features required for acceptability of Existing Engineering Equivalency Evaluations (Section 2.2.7)" R -Risk Criteria:

Systems / Features required to meet the Risk Criteria for the Performance-Based Approach (Section 4.2.4)* 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.Risk criteria information provided in Table 4-3 originates from FC07823. All other information provided in Table 4-3 originates from EA10-044 (NEI 04-02 Table B-3).4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase Planned modifications, studies, and evaluations to comply with NFPA 805 are described in Attachment S.The plant configuration as reflected by Revision 11 of the FCS internal events PRA was used as the starting point for the fire PRA. The fire PRA includes credit for proposed plant and procedural modifications identified in Attachment W, and in Attachment S, Tables S-2 and S-3.The NFPA 805 NSPC analyses include credit for proposed plant and procedural modifications identified in Attachments A and C, and in Attachment S, Tables S-2 and S-3.Planned plant modifications not included in Revision 11 of the internal events PRA and the fire PRA, and the NFPA 805 NSPC analyses have been reviewed for impact on the NFPA 805 application.

Based on engineering judgment the modifications do not significantly alter the fire PRA and/or NFPA 805 NSPC analyses results. Based on engineering judgment the modifications do not change the conclusions of the fire risk evaluations.

These plant modifications are identified in recommendation REC-134, Attachment S, Table S-3, and will be formally incorporated into the fire PRA model and NFPA 805 NSPC analyses.

Any impact on the NFPA 805 application will be assessed, as part of LAR implementation. (See Attachment S).Page 59 Omaha Public Power District FCS NFPA 805 Transition Report 4.8.3 Supplemental Information

-Other Licensee Specific Issues 4.8.3.1 None Page 60 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 005 RequIred far: Required Fire Protection Fi, Ar Deciption Regltory Baiss Tyo S L E Rt Ft and System Details 01 Safety Injection and Containment 4.2.3.2 Spray Pump Area I Detection I N N N N N Feature None- -----Suppression None 02 Sae Injection an Containment 4.2.3.2 Spray Pumnp Area..11 Detection I N N Y N N Feature None Suppression None 03 Spent Regenerant Tank & Pump Area 4.2.3.2 Detection I N N Y N N Feature None Suppression None 06-1 Gas Decay Tank Area 4.2.3.2 Detection None Feature None Suppression None 06-2 Gas Compressor Area 4.2.3.2 Detection None Feature None Suppression None 06-3 Basement & Personnel Conidor Area I 4.2.3.2 Purmer/Waste Fiter Room Detection I Y N Y N N For hatchway area water curtain actuation Detection I Y N Y N N For stairway area water curtain actuation Detection I Y N Y N N General area coverage Feature None Suppression D Y N Y N N Corridor stairway and hatch area water curtain Page 61 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 605 Required for Required Fire Protection Fire Area Description eulatory Bass Tye IS II E R Fau ad Syst Details 06-4 Rctste Monitor Tank Area 4.2.32 Detection None Feature None Suppression None 0- Shutdown Heat Exchanger Area 1 4.2.3.2 Detection None Feature None Suppression None 0" Shutdown Heat Exchanger Area It 4.2.3.2 Detection None Feature None Suppression None 06-7 Letdown Heat Exchaner Area III 4.2.32 Detection I N N N N N Feature None Suppression None 06 Heat Exchanger and Pump Area 4.2.3.2 Detection I N N N N N Feature None Suppression None 08 Safety Injection and Refueling Water 42.3.2 (SIRW) Storage Tank Area Detection None Feature None Suppression None 09 Valve Area I 4.2.3.2 Detection I N N N N N Feature None Suppression None Page 62 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NPPA 805 Required for. Required Fire Protection Fi Ar Delpton RapigityOwas TWIOI II LE R I I III I II a S tm DetalIs 10 Charging Pump Area 42.3.2 Detection I N N N N N Feature None Suppression None 13 Mechanical Penetration Area 4.2.3.2 Detection I N N N N N Feature None Suppression None 16 Valve Area I! 42.3.2 Detection None Feature None Suppression None 19 Personnel Complex Area 42.3.2 Detection I N N Y N N Feature None Suppression None 20-1 Personnel Corridor Area 4.2.4.2 Detection I Y N Y N Y Main floor corridor and personnel air lock Feature None Suppression None 20-2 Waste Evaporator Area 42.3.2 Detection None Feature None Suppression None 20-3 Volume Control Tank Area 4.2.3.2 Detection None Feature None Suppression None Page 63 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 805 Required for Required Fire Protection Fir At" Dewsction Regulatory Basi Type S L E R 0 Feature and System Details 20-4 Valve Area Iii 42.3.2 Detection None Feature None Suppression None 20-5 Ion Exchange Area 42.3.2 Detection None Feature None Suppression None 20-6 Waste Disposal and Hot Tool Storage 4.2.3.2 and Issue Area / Spent Resin Pump and Tank Room Detection I N N N N N Waste disposal and hot tool storage and issue area Feature None Suppression None 20-7 New Fuel Storage and Uncrating 4.2.3.2 Waste Hoklup Tank Area / Transfer Canal Pump / Ventilation Room Detection I N N Y N N Transfer canal pump room Detection I N N Y N N New fuel storage and uncrating Detection I N N Y N N Duct smoke detection Detection BT N N Y N N Loading and unloading area / spent fuel area projected beam detectors Detection I N N Y N N Ventilation room Feature None Suppression None 20-7ROOF VA-46A and VA-468 Condenser Area 42.4.2 Detection None Feature None Suppression None Page 64 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 805 Required for: Required Fire Protection Fire Area Descrption Regulatory Basis Type 8 L E Rt 0 Fetr n YSytem Details 23 Pipe Penetration Area 4.2.3.2 Detection I N N N N N Feature None Suppression None- -----24 Sampling Area 4.2.3.2 Detection I N N N N N Feature None Suppression None 28 Ventlation Equipment Area 4.2.4.2 Detection I N N Y N Y Feature None Suppression None 30 Containment 4.2.3.2 Detection LT Y Y Y N Y Cable tray Protectowire Detection T Y Y Y N Y Containment general area and reactor coolant pump cubicles Feature RES Y Y N N N Suppression None 31 Intake Structu 4.2.4.2 Detection I Y Y Y N Y Detection T Y N Y N Y Feature IFAB N Y N N N Pyrocrete enclosure protecting conduits associated with RW pumps Feature RES N Y N N N Concrete wall protecting RW pumps Feature ERFBS Y Y N N N Pyrocrete enclosure at ductbank entry into intake structure Suppression None Page 65 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 805 Required for. Required Fire Protection Fire Area Desciption Regulatory Sealsi T70 -S L I. R Fetr an System~ Details 32 Compressor Area 4.2...2 Detection I Y Y Y Y Y For actuation of compressor area pre-action sprinkler system Detection I N N Y N N Corridor to west switchgear room Feature RES N N N N Y Cable tray bottoms in overhead of room 19 Feature RES Y Y N N Y "V-shaped barrier protecting AFW pumps Suppression PA Y Y Y Y Y compressor area Suppression WP N N N N N AFW pump FW-10 partial system 33 Component Cooling Hag Exchanger 4.2.3.2 Area Detection I N N Y N N Feature None Suppression None 34A Eletrical Penetration Area -Basement 4.2.4.2 Detection I Y Y Y N Y Feature ERFBS Y Y N N Y 3M Interam wrap for train A MCC power cables Suppression None 348-1 Elecbcal Penetration Area Ground 4.2.4.2 and Intermediate Levels / QA Vault Detection I N N Y N N For actuation of QA vault halon system Detection I N N Y N Y Electrical penetration area ground and intermediate levels Feature None Suppression H N N Y N N QAvault 34C Group I MCC Area 4.2.3.2 Detection I N N N N N Feature None Suppression None Page 66 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 805 Required for: Required Fire Protecton Fire Ame Description Regulatory

&lsI TYPe S L E R~ ) Feature and System Details 35A Diesel Generator Room 1 4.2.3.2 Detection I N N Y Y N General area coverage and duct smoke detection Detection UV N N Y N N Diesel generator area 1 Detection P N N Y N N Diesel generator area 1 Feature None Suppression DP N N Y Y N 358 Diesel Generator Room 2 4.2.32 Detection P N N Y N N Diesel generator area 2 Detection UV N N Y N N Diesel generator area 2 Detection I N N Y Y N General area coverage and duct smoke detection Detection T N N Y N N Diesel ventilation enclosure Feature None Suppression DP N N Y Y N 36A East Switcear Area 4.2.4.2 Detection I Y Y N Y Y For halon system actuation Feature ERFBS Y N N N Y Pyrocrete enclosures protecting vertical cable trays and horizontal condut bank Feature IFAB Y Y N N Y Concrete block enclosure protecting panel AI-1 09B Suppression H Y Y N Y Y East/west switchgear area 368 West SLt Area (301) 4.2.4.2 Detection I N Y Y Y Y For halon system actuation Feature None Suppression H N Y Y Y Y East/west switchgear area 36C West Swkcgw Area (3MC) 4.2.3.2 Detection None Feature None Suppression None Page 67 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NiPA 605 Required for: Required Fire Protection FieArea DoscAonIII Rsgulatory B"Isl TYPe 8 L E R 0Feature and System Detail 37 Battery Room 1 4.2.3.2 Detection I N N N N N Duct smoke detection Feature None- -----Suppression None 38 Btery Room 2 4.2.3.2 Detection I N N N N N Duct smoke detection Feature None Suppression None 40 Equiment Hatch Enclosur Area 4.2.3.2 Detection None Feature None Suppression None 41 Cable Spreading Room 4.2.4.2 Detection I Y N Y Y Y For halon system actuation Feature None Suppression H Y N Y Y Y 42 Cont Room Complex Area 4.2.4.2 Detection I Y Y Y Y Y Mechanical equipment room Detection I Y Y Y N Y General area coverage and halon systern actuation Feature None Suppression H Y N Y N Y Main control room cabinet 43 Service and Emergency Feedwater 4.2.4.2 Storage Tank Area Detection IN N N Y N Y Air sampling system Detection I N N Y N Y Duct smoke detection Feature None Suppression MS N N N N N Control room HVAC charcoal filter manual water spray Page 68 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 605 Required for Required Fire Protection FirArea Description Regudetory RBees TyPe S L E R D Feature and System Deteils 45 Service Building 42.3.2 Detection I N N N N N Detection I N N N N N Duct smoke detection Detection T N N N N N Detection P N N N N N Feature None Suppression WP N N N N N System engineering office, administration offices, NRC offices, and programs engineering 46 Turbine Building I South Air Lock 4.2.3.2 Control Room to Turbine Building Detection UV/IR N N N N N Diesel driven AFW pump room Detection BT N N N N N Operating level projected beam detectors Feature None Suppression PA N N N N N Turbine generator bearings Suppression C N N N N N Turbine generator exciter housing Suppression WP N N N Y N Diesel driven AFW pump room Suppression WP N N N Y N Basement, mezzanine, and 1000' FW-16 heater platform Suppression WP N N N Y N Operating level office area Suppression D N N Y N N Water curtains protecting of control roorn entry doors Page 69 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 805 Required for Required Fire Protection FiA"De iptioh Regulatory 5a51 T yPe 8 L E R F e atue nd ystem Details 47 Transformer Yard ra Including 4.2.3.2 161/4kV Transformers TIA-3 and TIA-4 and their Disconnect Switches.

DS-TIA-3 and DS-TIA-4, the Overhed 161kV Transmission Lines from TlA-3 and TIA-4 to the 161kV Substation

1. 1251, the 4kV Bus Ducts from TIA-3 and TIA-4 into the Auxiiay Building$4 Swigear Rooms] and Breakers 1251-110 and 1251-111 at the 161kV Substation

1. 1251; Including Terminal Boxes PB-127T. PB-128T, and PB-1 29T Including the Manholes and Underground Ducts for the RW System Cables from the Auxilary Builing to the Intake Stucture;Including the Manholes and Underground Ducts for the 161kV Substation

1. 1251 Cables from the Auxdilary Building to the Substation

1. 1251 Control Building In the Switchyar Detection T N N N Y N Transformer deluge water spray detection system Feature None Suppression WS N N N Y N Transformer deluge water spray system 50 Outoor Ga Storage 4.2.3.2 Detection None Feature None Suppression None Page 70 Omaha Public Power District FCS NFPA 805 Transition Report TABLE 4.3 -

SUMMARY

OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES NFPA 605 Required for: Required Fire Prowtecton Fire Aree Description Regulatory Buis Type 8 L E R Feature and Sysema Details O/RP Chemistry and Radiation rection 4.2.32 Faclity Detection I N N N N N Under computer room floor Detection T N N N N N Detection I N N N N N Duct smoke detection Feature None Suppression WP N N N N N RW Radwast. Procesung Building 4.2.3.2 Detection I N N N N N Detection T N N N N N Detection I N N N N N Duct smoke detection Feature None Suppression WP N N N N N TSC Technical Support Center 4.2.3.2 Detection I N N N N N Detection I N N N N N Duct smoke detection Feature None Suppression D N N N N N Protection of ventilation unit VA-i119 YO Yard Ae, Condensate Storap 4.2.3.2 Tanku Genral Detection None Feature None Suppression None Page 71 Omaha Public Power District FCS NFPA 805 Transition Report Legend:___ ___ ___Type:,' Detection

' SuppreSsion Feature BT -Beam-Type C -Carbon Dioxide ERFBS -Etectical Raceway Fire Barrier System I Ionization CA -Clean Agent IFAB -Intra-Fire Area Barrier IN -Incipient F -Foam RES -Radiant Energy Shield IR Infrared D -Deluge LT -Line-Type DC -Dry Chemical P -Photoelectric DP -Dry Pipe T :-Thermal H -Halon UV ' -Ultravioet MS -Manual Spray UV/IR .UV/IR Combination PA -Pre-action WM -Water Mist WP -Wet Pipe WS -Water Spray-..-7 Required for:. -S -Required for Chapter 4 Separation Criteria L, -Required for NRC-Approved Licensing Action E -Required for Existing Engineering Equivalency Evaluation R -Required for Risk Significance P -Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or Fire Risk Evaluation Page 72 Omaha Public Power District FCS NFPA 805 Transition Report

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 FP requirements.

10 CFR 50.48 endorses, with exceptions, the NFPA's 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 OPPD does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, GDC 3, Fire Protection.

The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference 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(f).

Those regulatory requirements continue to apply to licensees that adopt NFPA 805. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirements may be met is different than under 10 CFR 50.48(b).

Specifically, whereas GDC 3 refers to SSCs important to safety, NFPA 805 identifies fire protection systems and features required to meet the Chapter 1 performance criteria through the methodology in Chapter 4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii) requirement to limit fire damage to SSCs important to safety so that the capability to safely shut down the plant is 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 nuclear safety performance criteria in Section 1.5 of NFPA 805. Once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design must meet any applicable requirements of NFPA 805, Chapter 3.Having identified the required fire protection systems and features, the licensee selects either a deterministic or performance-based approach to 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 probability and 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 CFR50.48(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 Page 73 Omaha Public Power District FCS NFPA 805 Transition Report the licensee's decision to comply with NFPA 805." Therefore, to the extent that the contents of the existing FP program plan required by 10 CFR 50.48(a) are inconsistent with NFPA 805, the FP 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 (ML081300697).

The following tables provide a cross reference of FP regulations associated with the post-transition FCS FP program and applicable industry and FCS documents that address the topic.10 CFR 50.48(a)Table 5-1 10 CFR 50.48(a) -Applicability/Compliance Reference 10 CFR 50.48(a) Section(s)

Applicability/Compliance Reference (1) Each holder of an operating license issued under See below this 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 these NFPA 805 Section 3.2.2 positions to implement those responsibilities; and NEI 04-02 Table B-1 (iv.) Outline the plans for fire protection, fire detection and NFPA 805 Section 2.7 and Chapters 3 and 4 suppression capability, and limitation of fire damage. NEI 04-02 B-1 and B-3 Tables (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 for NFPA 805 Sections 3.3.1 and 3.4 fire prevention and manual fire suppression activities; NEI 04-02 Table B-1 (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, systems, NFPA 805 Section 3.3 and Chapter 4 or components important to safety so that the capability to NEI 04-02 B-3 Table shut down the plant safely is ensured.Page 74 Omaha Public Power District FCS NFPA 805 Transition Report Table 5-1 10 CFR 50.48(a) -Applicability/Compliance Reference 10 CFR 50.48(a) Section(s)

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

[FCS Quality Assurance Plan QAP-3.4 Records Management and FCS Standing Order SO-C-2 FCS QualityAssurance Records](4) Each applicant for a design approval, design Not applicable.

FCS is licensed under 10 CFR 50.certification, or manufacturing license under part 52 of this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of Appendix A to this part.General Design Criterion 3 Table 5-2 GDC 3 -Applicability/Compliance Reference GDC 3, Fire Protection, Statement Applicability/Compliance Reference Structures, systems, and components important to NFPA 805 Chapters 3 and 4 safety shall be designed and located to minimize, NEI 04-02 B-1 and B-3 Tables consistent with other safety requirements, the probability and effect of fires and explosions.

Noncombustible and heat resistant materials shall be NFPA 805 Sections 3.3.2, 3.3.3, 3.3.4, 3.11.4 used wherever practical throughout the unit, NEI 04-02 B-1 Table particularly in locations such as the containment and control room.Fire detection and fighting systems of appropriate NFPA 805 Chapters 3 and 4 capacity and capability shall be provided and designed NEI 04-02 B-1 and B-3 Tables to minimize the adverse effects of fires on structures, systems, and components important to safety.Firefighting systems shall be designed to assure that NFPA 805 Sections 3.4 through 3.10 and 4.2.1 their rupture or inadvertent operation does not NEI 04-02 Table B-3 significantly impair the safety capability of these structures, systems, and components.

Page 75 Omaha Public Power District FCS NFPA 805 Transition Report 10 CFR 50.48(c)Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference Applicability/Compliance 10 CFR 50.48(c) Section(s)

Reference (1) Approval of incorporation by reference.

National Fire Protection Association General Information.

NFPA (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for 805 (2001 edition) is the Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), edition used.which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a)and 1 CFR 51.(2) Exceptions, modifications, and supplementation of NFPA 805. As used in General Information.

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

edition used.(i) Life Safety Goal, Objectives, and Criteria.

The Life Safety Goal, Objectives, The Life Safety Goal, and Criteria of Chapter 1 are not endorsed.

Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.(ii) Plant Damage/Business Interruption 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 performance Feed and bleed is not utilized criteria of Sections 1.5.1(b) and (c), a high-pressure charging/injection pump as the sole fire-protected safe coupled with the pressurizer power-operated relief valves (PORVs) as the shutdown methodology.

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 with Uncertainty analysis was not Section 2.7.3.5 is not required to support deterministic approach calculations.

performed for deterministic methodology.(v) Existing cables. In lieu of installing cables meeting flame propagation tests Electrical cable construction as required by Section 3.3.5.3, a flame-retardant coating may be applied to complies with a flame the electric cables, or an automatic fixed fire suppression system may be propagation test that was installed to provide an equivalent level of protection.

In addition, the italicized found acceptable to the NRC exception to Section 3.3.5.3 is not endorsed.

as documented in NEI 04-02 Table B-I.(vi) Water supply and distribution.

The italicized exception to Section 3.6.4 is not FCS 'Complies by Previous endorsed.

Licensees who wish to use the exception to Section 3.6.4 must NRC Approval" as submit a request for a license amendment in accordance with paragraph documented in NEI 04-02 (c)(2)(vii) of this section. Table B-1.Page 76 Omaha Public Power District FCS NFPA 805 Transition Report Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference Applicability/Compliance 10 CFR 50.48(c) Section(s)

Reference (vii) Performance-based methods. Notwithstanding the prohibition in Section 3.1 The use of performance-against the use of performance-based methods, the fire protection program based methods for NFPA 805 elements and minimum design requirements of Chapter 3 may be subject to Chapter 3 is requested.

See the performance-based methods permitted elsewhere in the standard.

Attachment L.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 NFPA The LAR was submitted in 805 as an alternative to complying with paragraph (b) of this section for accordance with 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.Page 77 Omaha Public Power District FCS NFPA 805 Transition Report Table 5-3 10 CFR 50.48(c) -Applicability/Compliance Reference Applicability/Compliance 10 CFR 50.48(c) Section(s)

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

were designee of the Director, may approve the application if the Director or utilized.designee determines that the proposed alternatives: (i) Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;(ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

5.2 Regulatory

Topics 5.2.1 License Condition Changes The current FCS fire protection license condition 3.D is being replaced with the standard license condition in Regulatory Position C.3.1 of RG 1.205, as shown in Attachment M.5.2.2 Technical Specifications OPPD conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded.

OPPD determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the OPPD adoption of the new FP licensing basis.5.2.3 Orders and Exemptions A review was conducted of the OPPD docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained.

A discussion of affected orders and exemptions is included in Attachment

0.5.3 Regulatory

Evaluations 5.3.1 -No Significant Hazards Consideration A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:* Involve a significant increase in the probability or consequences of an accident previously evaluated; or Page 78 Omaha Public Power District FCS NFPA 805 Transition Report" 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. OPPD has evaluated the proposed amendment and determined that it involves no significant hazards consideration.

5.3.2 Environmental

Consideration Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c).

That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.

5.4 Transition

Implementation Schedule The following schedule for transitioning FCS to the new FP licensing basis requires NRC approval of the LAR in accordance with the following schedule: " Implementation of new NFPA 805 FP program to include procedure changes, process updates, and training to affected plant personnel.

This will occur six (6)months after NRC approval." OPPD will complete modifications necessary to support the new FP licensing basis for transitioning to NFPA 805 by the end of the second Refueling Outage following NRC approval.

Appropriate compensatory measures will be maintained until modifications are complete.Page 79 Omaha Public Power District FCS NFPA 805 Transition Report

6.0 REFERENCES

The following references were used in the development of the TR.Regulatory Guidance 6.1. NUREG-1805, Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U. S. Nuclear Regulatory Commission Fire Protection Inspection Program, US NRC -Office of Nuclear Reactor Regulation, December 2004 6.2. NUREG-1824 and EPRI 1011999, Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 7: Fire Dynamics Simulator (EPRI 1011999), US NRC -Office of Nuclear Regulatory Research, May 2007 6.3. NUREG-1921 and EPRI TR-1019196, Draft Report for Public Comment, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, EPRI, Palo Alto, CA, and USNRC, Office of Nuclear Regulatory Research, Washington, D.C.: November 2009 6.4. NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Electric Power Research Institute

-EPRI 1011989, Final Report, September 2005 6.5. NUREG/CR-6850 Supplement 1, Fire Probabilistic Risk Assessment Methods Enhancements, Supplement 1 to NUREG/CR-6850 and EPRI 1011989, Electric Power Research Institute

-EPRI 1019259, Technical Report, September 2010 6.6. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1 -November 2002 6.7. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 -March 2009 6.8. Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Revision 1 -December 2009 Codes and Standards 6.9. American Society for Testing and Materials (ASTM) ASTM-E-1 19-73, Fire Tests of Building Construction and Materials 6.10. American Society for Testing and Materials (ASTM) ASTM-E-84-50T, Test Method for Surface Burning Characteristics of Building Materials 6.11. American Society for Testing and Materials (ASTM) ASTM E-648, Standard Test Method for Critical Radiant Flux of Floor-Covering Systems Using a Radiant Heat Source Page 80 Omaha Public Power District FCS NFPA 805 Transition Report 6.12. ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, December 2009 6.13. ASME RA-Sb-2005, Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, December 2005 6.14. ASME RA-Sa-2003, Addenda to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, American Society of Mechanical Engineers, New York, NY, October 2003 6.15. IEEE 384, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits, 1977 Edition 6.16. IEEE-383, IEEE Standard for Type Test of Class 1 E Electric Cables, Field Splices, and Connections Nuclear Power Generating Stations, 1977 Edition 6.17. NFPA 10, Standard for Portable Fire Extinguishers, 1994 Edition 6.18. NFPA 101, Code for Safety to Life from Fire in Buildings and Structures, 2000 Edition 6.19. NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 1993 Edition 6.20. NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems, 1992 Edition 6.21. NFPA 13, Standard for the Installation of Sprinkler Systems, 1996 Edition 6.22. NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1996 Edition 6.23. NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1996 Edition 6.24. NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, 1984 Edition 6.25. NFPA 30, Flammable and Combustible Liquids Code, 1966, 1987, and 2000, 2008 Editions 6.26. NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, 1969 Edition 6.27. NFPA 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, 1999 Edition 6.28. NFPA 72, National Fire Alarm Code, 1990, 2002, and 2007 Editions 6.29. NFPA 72D, Standard for the Installation, Maintenance and Use of Proprietary Protective Signaling Systems, 1975 and 1979 Editions 6.30. NFPA 72E, Standard on Automatic Fire Detectors, 1978, 1982, 1990 Editions 6.31. NFPA 80, Standard for Fire Doors and Windows, 1968 Edition Page 81 Omaha Public Power District FCS NFPA 805 Transition Report 6.32. NFPA 80A, Recommended Practice for Protection of Buildings from Exterior Fire Exposures, 1996 Edition 6.33. NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, 1985 Edition 6.34. NFPA 220, Standard on Types of Building Construction, 1999 Edition 6.35. NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations, 2000 Edition 6.36. NFPA 251, Standard Methods of Tests of Fire Endurance of Building Construction and Materials, 1999 Edition 6.37. NFPA 252, Standard Methods of Fire Tests of Door Assemblies, 1999 Edition 6.38. NFPA 256, Standard Methods of Fire Tests of Roof Coverings, 1998 Edition 6.39. NFPA 600, Standard on Industrial Fire Brigades, 2000 Edition 6.40. NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, 1999 Edition 6.41. NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition Correspondence 6.42. Letter from OPPD (Theodore E. Short) to NRC (George Lear), dated December 30, 1976 (LIC-76-0180) 6.43. Letter from OPPD (T. E. Short) to NRC (G. E. Lear), "Application for Amendment of Operating License for the Fort Calhoun Station, Unit 1," dated February 16, 1977 (LIC-77-0172) 6.44. Letter from OPPD (T. E. Short) to NRC (G. E. Lear), "The District's Response to Questions from NRC Regarding Fire Protection System at Fort Calhoun Responses to Questions 13,14,17,19

& Staff (44252)" dated September 17,1977 (LIC-77-0103) 6.45. Letter from NRC (G. Lear) to OPPD (T. E. Short), dated February 14, 1978 (NRC-78-0025) 6.46. Letter from NRC (R. W. Reid) to OPPD (T. E. Short), dated August 23, 1978 (NRC-78-0104) 6.47. Letter from OPPD (T. E. Short) to NRC (R. W. Reid), "Application for Amendment of Operating License Request for the Fort Calhoun Station, Unit 1," dated January 22, 1979 (LIC-79-0171) 6.48. Letter from OPPD (T. E. Short) to NRC (R. W. Reid), "Result of a conference call held on May 23, 1979 between the Omaha Public Power District and members of the Commission's staff', dated July 9, 1979 (LIC-79-0192) 6.49. Letter from NRC (R. A. Clark) to OPPD (W. C. Jones), dated November 17, 1980 (NRC-80-0213)

Page 82 Omaha Public Power District FCS NFPA 805 Transition Report 6.50. Letter from OPPD (W.C. Jones) to NRC (R. A. Clark), "Response to NRC Questions -Post Fire Safe Shutdown Capability Fort Calhoun Nuclear Power Station," dated March 27, 1981 (LIC-81-0042) 6.51. Letter from NRC (T. M. Novak) to OPPD (W. C. Jones), "Fire Protection Rule-10 CFR 50.48(C)(5)

Alternative Safe Shutdown-Section III.G.3 of Appendix R to 10CFR50, Fort Calhoun Nuclear Power Station," dated April 8, 1982 (NRC-82-0060) 6.52. Letter from NRC (R. A. Clark) to OPPD (W. C. Jones), "Fire Protection Rule," dated August 12, 1982 (NRC-82-0149) 6.53. Letter from NRC (W. C. Seidle) to OPPD (W. C. Jones), dated July 1, 1983 (NRC-83-0202) 6.54. Letter from OPPD (W. C. Jones) to NRC (R. A. Clark), "Request for Exemptions from Various Requirements of 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities," dated August 30, 1983 (LIC-83-0219) 6.55. Letter from OPPD (R. L. Andrews) to NRC (J. R. Miller) "10 CFR50, Appendix R," dated December 3, 1984 (LIC-84-041 1)6.56. Letter from OPPD (R. L. Andrews) to NRC (J. R. Miller), "1 OCFR50 Appendix R Exemption Request Revisions," dated January 9, 1985 (LIC-84-0338) 6.57. Letter from OPPD (R. L. Andrews) to NRC (J. R. Miller) "10 CFR50, Appendix R," dated March 3, 1985 (LIC-85-0037) 6.58. Letter from NRC (E. J. Butcher) to OPPD (R. L. Andrews), "Exemption Requests for the Fort Calhoun Station, Unit No. 1 1OCFR Part 50, Appendix R. Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," dated July 3, 1985 (NRC-85-0200) 6.59. Letter from NRC (E. J. Butcher) to OPPD (R. L. Andrews), "Alternate Shutdown Capability for the Upper Electrical Penetration Room," dated November 5,1985 (NRC-85-0306) 6.60. Letter from OPPD (R. L. Andrews) to NRC (A. C. Thadani), "10 CFR 50 -Appendix R," dated April 9, 1986 (LIC-86-0118) 6.61. Letter from NRC (D. E. Sells) to OPPD (R. L. Andrews), "Appendix R, 1 OCFR PART 50, Fire Protection Modifications," dated July 1, 1986 (NRC-86-021 1)6.62. Letter from NRC (D. E. Sells) to OPPD (R. L. Andrews), "Control Room Carpet," dated December 22, 1986 (NRC-86-0408) 6.63. Letter from OPPD (K. J. Morris) to NRC (Document Control Desk),'Exemption Request from 10 CFR 50 Appendix R, Section 111.0, "Oil Collection System for Reactor Coolant Pumps",' dated November 28, 1988 (LIC-88-1066)Page 83 Omaha Public Power District FCS NFPA 805 Transition Report 6.64. Letter from NRC (J. A. Calvo) to OPPD (K. J. Morris), "Fort Calhoun Station, Unit 1 -Exemption from the Requirements of Appendix R to 10CFR PART 50 (Section 111.0)," dated December 20, 1988 (NRC-88-0457) 6.65. Letter from NRC (D. M. Crutchfield) to OPPD (W. G. Gates), "Denial of Exemption Request for Fire Area 34B -Fort Calhoun Station, Unit 1 (TAC NO. 76055)," dated November 14, 1990 (NRC-90-0379) 6.66. Letter from OPPD (W. G. Gates) to NRC (Document Control Desk),"Clarification of 10CFR Part 50, Appendix R Exemption Request for Fort Calhoun Station (FCS) Fire Area 34A (Lower Electrical Penetration Area)," dated August 5, 1992 (LIC-92-217R) 6.67. Letter from NRC (S. Bloom) to OPPD (T. L. Patterson), "Clarification of 1OCFR Part 50, Appendix R Exemption Request for Fort Calhoun Station (FCS) Fire Area 34A (Lower Electrical Penetration Area) (TAC NO. M84457)," dated March 17, 1993 (NRC-93-0095) 6.68. Letter from OPPD (W. G. Gates) to NRC (Document Control Desk), "NRC Generic Letter 88720 Submittal for Fort Calhoun Station "Individual Plant Examination for Severe Accident Vulnerabilities" (TAC No. 74412)," dated December 1, 1993 (LIC-93-0278) 6.69. Letter from NRC (S. Bloom) to OPPD (T. L. Patterson), "Fort Calhoun Station, Unit No. 1 -Amendment No. 160 to Facility Operating License No. DPR-40 (TAC NO. M87825)," dated January 14, 1994 (NRC-94-0015) 6.70. Letter from OPPD (T. L. Patterson) to NRC (Document Control Desk)," Phase II Response to Generic Letter 88-20, Supplement 4 Individual Plant Examination of External Events (IPEEE)," dated June 30, 1995 (LIC-95-0130) 6.71. Letter from NRC (L. R. Wharton) to OPPD (T. L. Patterson), "Fort Calhoun Station, Unit No. 1 -Review Of Individual Plant Examination (IPE) Submittal

-Internal Events (TAC NO. M74412)," dated December 9, 1996 (NRC-96-0216)6.72. Letter from OPPD (S. K. Gambhir) to NRýC (Document Control Desk),"Response to Questions Related to 10 CFR Part 50, Appendix R as proposed in NRC Inspection Report 50-285/96-16," dated February 7, 1997 (LIC-97-0015)6.73. Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk),"Request for Exemption from Requirements of 1OCFR50 Appendix R and Licensing Basis Update -Reactor Coolant Pump Motor Lube Oil Collection System," dated September 30, 1997 (LIC-97-0073) 6.74. Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk),"Additional Information to Support Request for Exemption from 1 OCFR50 Appendix R. and Updated Status on Reactor Coolant Pump Motor Lube Oil Corrective Action," dated January 29, 1998 (LIC-98-0001) 6.75. Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk),"Clarification to Request for Exemption from Requirements of 1 OCFR50 Page 84 Omaha Public Power District FCS NFPA 805 Transition Report Appendix R -Reactor Coolant Pump Motor Lube Oil Collection System," dated April 23, 1998 (LIC-98-0063) 6.76. Letter from NRC (L. R. Wharton) to OPPD (S. K. Gambhir), "Issuance of Exemption from the Requirements of 10CFR Part 50. Appendix R, Fort Calhoun Station, Unit No.1 (TAC NO. M99724)," dated May 21, 1998 (NRC-98-0083)6.77. Letter from OPPD (H. J. Faulhaber) to NRC (Document Control Desk), 'Fort Calhoun Station Unit No.1, Supplemental Response to Generic Letter 2006-03, "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations",'

dated August 2, 2006 (LIC-06-0076) 6.78. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure), "Fort Calhoun Station, Unit No. 1 -Conforming License Amendment to Incorporate the Mitigation Strategies Required by Section B.5.b of Commission Order EA-02-026 (TAC No. MD4534)" dated July 26, 2007 (NRC-07-0073) 6.79. Letter from OPPD (R.P. Clemens) to NRC (Document Control Desk),"Request for Exemption from Requirements of 10 CFR 50, Appendix R, Section III.G.1 .b for Fire Area 31 at the Fort Calhoun Station," dated February 4, 2008 (LIC-08-0006) 6.80. Letter from OPPD (D. J. Bannister) to NRC (Document Control Desk), "Letter of Intent to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition," dated June 9, 2008 (LIC-08-0069) 6.81. Letter from OPPD (R. P. Clemens) to NRC (Document Control Desk),"Response to Request for Additional Information Concerning Exemption from Requirements of 10 CFR 50, Appendix R, Section III.G.1 .b for Fire Area 31 at the Fort Calhoun Station," dated October 31, 2008 (LIC-08-0109) 6.82. Letter from NRC (J. G. Giitter) to OPPD (D. J. Bannister), "Fort Calhoun Station, Unit No. 1 -Response to Letter of Intent to Adopt National Fire Protection Association Standard 805 (TAC NO. MD9736)," dated November 20, 2008 (NRC-08-0107) (ML082710002) 6.83. Letter from NRC (A. B. Wang) to OPPD (D. J. Bannister), "Fort Calhoun Station. Unit No.1 -Exemption from the Requirements of 1OCFR Part 50, Appendix R,Section III.G.1 .b (TAC NO. MD8049)," dated February 6, 2009 (NRC-09-0007) 6.84. CFTC-09-56, Letter from Westinghouse Electric Company (G. D. Auld) to OPPD (Joe McManis), "Omaha Public Power District Fort Calhoun Nuclear Station NFPA 805 Fire PRA, NFPA 805 -Task 4.6 Sample Cable/Raceway Fire Area Location," dated July 27, 2009 6.85. CFTC-09-85, Letter from Westinghouse Electric Company (G. D. Auld) to OPPD (J. McManis), "NFPA 805 -Task 1.1 Fire Protection Program Design/Licensing Document Mapping," datedOctober 8, 2009 Page 85 Omaha Public Power District FCS NFPA 805 Transition Report 6.86. CFTC-09-92, Letter from Westinghouse Electric Company (G. D. Auld) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA 805 -PIF-09-001, Task 1.3 -Qualitative Fire Risk Insights for EPU," dated October 30, 2009 6.87. Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk), "10 CRF 50.59 Report, Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, and Updated Safety Analysis Report (USAR)Revision for Fort Calhoun Station (FCS), Unit No. 1," dated June 18, 2010 (LIC-10-0040) 6.88. NFPA805-WEST-10-005, Letter from OPPD (J. McManis) to Westinghouse (G. Samide), "Cable Verification Task 4.7 to Address F&O CS-C2-01 ," December 14, 2010 6.89. CFTC-10-224, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 7.19 Fire Risk-Beneficial Modifications," dated December 17, 2010 6.90. NFPA805-WEST-1 1-001, Letter from OPPD (J. McManis) to Westinghouse (C. Worrell), "Cable FPRA Cable Sample Study to Address FPRA F&Os," January 7, 2011 6.91. CFTC-1 1-57, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 2.3 Final 86-10 Engineering Analyses," dated March 7, 2011 6.92. OG-1 1-83, Letter from PWR Owners Group (T. Zachariah) to OPPD Risk Management Subcommittee Representative (A. Hackerott), Transmittal of Letter Report LTR-RAM-II-10-072, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications For the Fort Calhoun Fire Probabilistic Risk Assessment," PA-RMSC-0403, Westinghouse Electric Company, dated March 7, 2011 6.93. CFTC-11-71, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 2.2 NFPA Code Compliance Review," dated March 17, 2011 6.94. CFTC-1 1-84, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 2.1 Chapter 3 Fundamental Fire Protection Program & Design Elements Review," dated March 25, 2011 6.95. CFTC-1 1-85, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 4.1 Nuclear Safety Performance Analysis Methodology Review," dated March 25, 2011 Page 86 Omaha Public Power District FCS NFPA 805 Transition Report 6.96. CFTC-1 1-95, Letter from Westinghouse Electric Company (C. M. Burton) to OPPD (J. McManis), "Omaha Public Power District, Fort Calhoun Station, NFPA 805 Fire PRA, NFPA-805 -Task 7.17 PRA Peer Review History," dated April 1, 2011 6.97. Letter from OPPD (H. J. Faulhaber) to NRC (Document Control Desk),"Notification of Extension of Commitment to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) at Fort Calhoun Station Unit No. 1," dated June 9, 2011 (LIC-11-0054)6.98. Letter from NRC (J. G. Giitter) to OPPD (D. J. Bannister), "Fort Calhoun Station, Unit No. 1 -Response to Request for Extending Enforcement Discretion Deadline to Adopt National Fire Protection Association Standard 805 (TAC No. ME6643)," dated July 22, 2011 (NRC-11-0079)(ML111960435)

Engineering Analyses, Calculations, and Design Changes 6.99. EA10-036, N FPA 805 Deterministic Safe Shutdown Separation Analysis, Revision 0 6.100. EA10-037, NFPA 805 NSPC and Fire PRA Circuit Analysis, Cable Selection, and Cable Location, Revision 0 6.101. EA10-041, Operator Manual Action Feasibility Assessment, Revision 0 6.102. EA10-042, Non-Power Operation Modes Transition Review, Revision 0 6.103. EA10-043, Radioactive Release Review, Revision 0 6.104. EA10-044, NFPA 805 Recovery Action Feasibility Assessment, Revision 0 6.105. EA10-045, NFPA 80 Fire Door Deviations, Revision 0 6.106. EA10-046, QA Vault Abort Switch , Revision 0 6.107. EA10-047, NFPA 12 C02 Deviations, Revision 0 6.108. EA10-048, NFPA 805 Lack of Suppression Intake Structure, Revision 0 6.109. EA10-060, Turbine Generator Exciter C02 -Lack of Odorizer, Revision 0 6.110. EA10-062, NFPA 805 Chapter 3 Fundamental Fire Protection Program and Design Elements Review, Revision 0 6.111. EA10-063, Engineering Evaluation Review, Revision 0 6.112. EA10-064, Nuclear Safety Performance Analysis Methodology Review, Revision 0 6.113. EA1 1-006, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 20-1, Revision 0 6.114. EA1 1-007, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 20-7ROOF, Revision 0 Page 87 Omaha Public Power District FCS NFPA 805 Transition Report 6.115. EA1 1-008, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 28, Revision 0 6.116. EA11-010, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 31, Revision 0 6.117. EA1 1-011, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 32, Revision 0 6.118. EA1 1-012, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 34A, Revision 0 6.119. EA11-013, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 341B-1, Revision 0 6.120. EA11-014, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 36A, Revision 0 6.121. EA 1-015, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 36B, Revision 0 6.122. EA1 1-016, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 41, Revision 0 6.123. EA11-017, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 42, Revision 0 6.124. EA11-018, Risk-Informed Performance Based Fire Risk Evaluation for Fire Area 43, Revision 0 6.125. EA-FC-06-005, Fire Barrier Evaluation Between Room 66 and Containment Stressing Gallery, Revision 0 6.126. EA-FC-89-055, 10 CFR 50 Appendix R Safe Shutdown Analysis, Revision 15 6.127. EA-FC-90-088, Fire Resistance Rating of Doors 1036-1, 1036-2 and 1011-28, Revision 0 6.128. EA-FC-91-019, Grouted Pen. Seal Qualification, Revision 3 6.129. EA-FC-91-020, Caulk and Fiber Pen. Seal Qualification, Revision 4 6.130. EA-FC-91-021, Boot Seal Qualification, Revision 3 6.131. EA-FC-91-022, 9" Silicone Foam Pen. Seal Qualification, Revision 6 6.132. EA-FC-91-023, 6" Silicone Foam Pen. Seal Qualification, Revision 5 6.133. EA-FC-91-084, Breaker/Fuse Coordination Study, Revision 7 6.134. EA-FC-92-014, Throughbolt Penetration Evaluation, Revision 0 6.135. EA-FC-92-017, Evaluation of Sprinkler System & Water Supply for Room 19, Revision 0 6.136. EA-FC-92-043, Evaluation of Fire Barrier Separating Rooms 25 and 26, Revision 1 Page 88 Omaha Public Power District FCS NFPA 805 Transition Report 6.137. EA-FC-93-033, Evaluation of Fire Barriers to GL 92-08 and Evaluation of Miscellaneous Fire Barriers, Revision 3 6.138. EA-FC-93-042, Containment Penetration Seal Evaluation, Revision 1 6.139. EA-FC-93-047, Halon System Operability Evaluation, Revision 3 6.140. EA-FC-94-032, Fire Pump Strainer Evaluation (FP-6A & FP-6B), Revision 2 6.141. EA-FC-94-039, Internal Sealing of Conduits, Revision 0 6.142. EA-FC-95-022, NFPA Code Compliance, Revision 3 6.143. EA-FC-96-015, Non-Safety Related Fire Barriers, Revision 3 6.144. EA-FC-97-001, Updated Fire Hazards Analysis, Revision 14 6.145. EA-FC-97-002, RCP Lube Oil Fire Hazard Evaluation, Revision 1 6.146. EA-FC-97-014, 86-10 Analysis of Auxiliary Building Water Curtain Separating Fire Areas 6 and 20, Revision 2 6.147. EA-FC-97-015, Fire Barrier Evaluation Between Room 22 and Containment Stressing Gallery, Revision 0 6.148. EA-FC-97-041, Analysis of Manual Hose Station Locations, Revision 0 6.149. EA-FC-97-043, Fire Safe Shutdown for Control Room Evacuation Design Basis Analysis, Revision 9 6.150. EA-FC-97-044, 10 CFR 50 Appendix R Cable Identification, Revision 9 6.151. EA-FC-98-001, Fire Barrier Evaluation for HVAC Penetrations, Revision 3 6.152. EA-FC-98-002, Fire Barrier Evaluation for GL 86-10 6" Foam Penetrations, Revision 5 6.153. EA-FC-98-003, GL 86-10 Fire Barrier Evaluation for 9 inch Foam Penetrations, Revision 8 6.154. EA-FC-98-004, Fire Barrier Evaluation for 86-10 Conduit Seals, Revision 2 6.155. EA-FC-98-005, Fire Barrier Evaluation for GL 86-10 Miscellaneous Penetrations, Revision 5 6.156. EA-FC-98-028, Fire Barrier Evaluation for Unmortared Block Walls (86-10), Revision 0 6.157. EA-FC-98-032, Fire Test Review EA, Revision 1 6.158. EA-FC-99-023, Fire Protection Suppression Effects Analysis, Revision 0 6.159. EC 28212, Replacement of Fire Detectors in Room 81, April 4, 2005 6.160. EC 31743, Upgrade Main Generator Protective Relays, January 23, 2003 6.161. EC 32183, Containment Fire Detection Upgrade, March 31, 2010 6.162. EC 32387, Turbine Controls System Replacement, April 25, 2003 Page 89 Omaha Public Power District FCS NFPA 805 Transition Report 6.163. EC 32591, Demin Water Storage Tanks for RO Unit Operations, May 30, 2003 6.164. EC 33663, Valve Positioner Replacements on E/P-101-1

& E/P-101-2, June 25, 2010 6.165. EC 35741, Traveling Screens Replacement, January 4, 2005 6.166. EC 37112, Automatic Blowdown Isolation, April 22, 2010 6.167. EC 37738, Fire Protection Piping Upgrade, January 26, 2010 6.168. EC 39000, Room 20 Fire Extinguisher, May 22, 2009 6.169. EC 40493, Installation of New Ball Valve FP-960, August 12, 2009 6.170. EC 40954, Maintenance Building Expansion, June 20, 2007 6.171. EC 41587, Raw Water Strainer Upgrade, October 10, 2007 6.172. EC 41956, DG Engine Controls Upgrade, September 13, 2007 6.173. EC 42682, Fire Hose Connections for FW-1229, April 7, 2009 6.174. EC 43209, FW-19 EFWST Expansion (EPU), May 1, 2008 6.175. EC 43758, Loop L106 Upgrade, Replace Transmitter, Manifold & Indicator, March 16, 2010 6.176. EC 44354, Fire Protection Main Line Modification, August 7, 2009 6.177. EC 44832, Enhance Emergency Lighting in Room 29 -VCT Room, November 18, 2008 6.178. EC 45105, FW-10 Speed Limiting Governor Setting, November 16, 2009 6.179. EC 46350, Install Operating Platform for AC-12A and AC-12B, November 15, 2010 6.180. EC 50655, FACTS Enhancements Identified During NFPA805 Upgrade, November 8, 2010 6.181. ECN-94-398, Screen House Manhole Conduit Seals, Revision 1 6.182. ECN-97-029, Modify RCP Motor Oil Collection System Drains, Revision 0 6.183. FC05188, Short Circuit Calculations for Coordination of Protective Devices on the Instrument Inverters, using EDSA Technical 2005, Revision 3 6.184. FC05814, UFHA Combustible Loading Calculation, Revision 11 6.185. FC06237, Room Heatup Due to Loss of HVAC, Revision 0 6.186. FC06355, 10 CFR50 Appendix R Functional Requirements and Component Selection, Revision 14 6.187. FC06672, Standpipe Hose Station Hydraulic Demand Calculation, Revision 0 6.188. FC06718, Appendix R -Miscellaneous Mechanical Support Systems Calculation, Revision 1 Page 90 Omaha Public Power District FCS NFPA 805 Transition Report 6.189. FC07083, Doses Due to Component Cover Plate Damage, Revision 0 6.190. FC07084, Original Steam Generator Storage Facility Flooding Concentrations, Revision 0 6.191. FC07818, Plant Boundary Definition and Partitioning, Westinghouse Electric Company, Revision 0 6.192. FC07819, Fire PRA Component Selection, Westinghouse Electric Company, Revision 0 6.193. FC07820, Fire Risk Model (CCDP & CLERP), Westinghouse Electric Company, Revision 0 6.194. FC07821, Fire Ignition Frequencies, Revision 0 6.195. FC07822, Fire Ignition Source Zone of Influence Determinations for Fort Calhoun Station, Westinghouse Electric Company, Revision 0 6.196. FC07823, Fire Scenario Selection and Characterization for Fort Calhoun Station, Westinghouse Electric Company, Revision 0 6.197. FC07824, Fire PRA Main Control Room Analysis, Revision 0 6.198. FC07825, Fire Human Reliability Analysis, Revision 0 6.199. FC07826, Qualitative Screening, Quantitative Screening, Quantification, and Uncertainty Analysis, Westinghouse Electric Company, Revision 0 6.200. FC07865, National Fire Protection Association (NFPA) Standard 805 Airborne and Liquid Effluents Offsite Dose, Revision 0 6.201. FC07869, NFPA 805 Recovery Actions Evaluation at FCS for EPU, Revision 0 6.202. FC07883, Fire Risk Assessment of FCS Variances from Deterministic Requirements of NFPA 805, Revision 0 Industry Documents/Software 6.203. EPRI 1003111, Fire Events Database and Generic Ignition Frequency Model for U.S. Nuclear Power Plants, EPRI, Palo Alto, CA, 2001 6.204. EPRI 1016735, Fire PRA Methods Enhancements:

Additions, Clarifications, and Refinements to EPRI 1019189, EPRI, Palo Alto, CA, December 2008.6.205. EPRI TR-100370, Fire-Induced Vulnerability Evaluation (FIVE), Electric Power Research Institute (EPRI), Palo Alto, CA, May 1992 6.206. EPRI TR-100443, Methods of Quantitative Fire Hazard Analysis, EPRI, Palo Alto, CA, May 1992 6.207. Fire Protection Handbook, 20th ed., National Fire Protection Association (NFPA), (2 Volumes), 2008 6.208. Fort Calhoun Safe Version 1.5.0 (SAFE-PB, EPM VER. 5.0.0)Page 91 Omaha Public Power District FCS NFPA 805 Transition Report 6.209. Karlsson, B., Quintiere, J. G., "Enclosure Fire Dynamics," CRC Press, 2000 6.210. Macleod, D., Kolonauski, L., and Grobbelaar, J., "Simplified HRA Process for Internal Fire Analysis," ANS PSA 2008 Topical Meeting, Knoxville, Tennessee, September 7-11, 2008 6.211. NEI 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, Revision 1, January 2005 6.212. NEI 00-02, Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, Revision A3, March 2000 6.213. NEI 04-02, Guidance for Implementing A Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2 6.214. NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 0, Draft H, November 2008 6.215. October 6,2010 Public Meeting Presentation "Generic Issue 199: Implications of Updated Probabilistic Seismic Hazard Estimates in the Central and Eastern United States on Existing Plants -Safety/Risk Assessment," USNRC (ML102770655) 6.216. Roadmap for Attaining Realism in Fire PRAs, prepared for: NEI Fire PRA Task Force, Nuclear Energy Institute (NEI), December 2010 (ML1 10210990)6.217. SFPE Handbook of Fire Protection Engineering, 4th ed., Society of Fire Protection Engineers, National Fire Protection Association (NFPA), 2008 6.218. WCAP-1 6933-NP, PWR Generic List of Fire-Induced Multiple Spurious Operation Scenarios, Revision 0, dated April 2009 Procedures, Manuals, Guidelines, and Forms 6.219. AOP-06, Fire Emergency, Revision 24 6.220. AOP-06-01, Fire Emergency, Auxiliary Building Radiation Controlled Areas and Containment, Revision 1 6.221. AOP-06-02, Fire Emergency, Uncontrolled Areas of Auxiliary Building, Revision 1 6.222. AOP-06-03, Fire Emergency, Miscellaneous Areas, Revision 0 6.223. AOP-13, Loss Of Control Room Air Conditioning, Revision 9 6.224. AOP-17, Loss Of Instrument Air, Revision 13 6.225. AOP-30, Emergency Fill Of Emergency Feedwater Storage Tank, Revision 9 6.226. ARP-CB10,1 1/Al1, B6-L Window, Annunciator Response Procedure Al1 Control Room Annunciator Al 1, Condensate Storage Tank LO, Revision 23 6.227. ARP-CB10, 11/A21, B5-L Window, Annunciator Response Procedure A21 Control Room Annunciator A21, Aux Feedwater Pump FW-54 Trouble, Revision 31 Page 92 Omaha Public Power District FCS NFPA 805 Transition Report 6.228. CH-ODCM-0001, Offsite Dose Calculation Manual (ODCM), Revision 20 6.229. EOP-00, Standard Post Trip Actions, Revision 27 6.230. EOP-20, Functional Recovery Procedure, Revision 24 6.231. EM-CP-06-FDZH-QA, Calibration Procedure

-Calibration and Functional Test of QA Vault Halon Fire Detection And Protection System, Revision 8 6.232. EM-CP-06-FDZ-TG, Calibration Procedure

-Disassembly, Reassembly and Testing Of Turbine Generator Fire Detector System, Revision 6 6.233. FC-18, Hot Work Permit, Revision 27 6.234. FC-1142, Fire Protection Impairment Permit, Revision 13 6.235. FCS Renewed Facility Operating License No. DPR-40, as of Amendment No.265 6.236. FCS PRA Summary Notebook, Revision 12 6.237. FCSG-2, Guideline

-Observation

-Quality Contact Hours Program, Revision 29 6.238. FCSG-1 5-11, Guideline

-Fire Prevention Plan, Revision 2 6.239. FCSG-15-24, Guideline

-Housekeeping, Revision 6 6.240. FCSG-15-33, Excavations, Trenches and Shoring, Revision 3 6.241. FCSG-15-35, Guideline-Welding, Cutting and Burning, Revision 6 6.242. FCSG-15-36, Guideline

-Compressed Gas Cylinder Safety, Revision 6 6.243. FCSG-33, Guideline

-Plant Health Committee Process, Revision 10 6.244. FCSG-56, Time Critical Operator Action Standard, Revision 0 6.245. GM-RM-FP-0303, Repetitive Maintenance

-Semi-Annual C02 Cylinders Inspection, Revision 6 6.246. IC-ST-FP-0001, Calibration and Functional Test of Auxiliary Building, Elevation 1036'(Room

81) Fire Detection System (AI-230 and AI-231), Revision 0 6.247. OI-FP-4, Fire Detection Systems, Revision 42 6.248. OI-RC-9, Reactor Coolant Pump Operation, Revision 73 6.249. OI-ST-6, Operating Instruction

-Operation of the Generator Gas System, Revision 27 6.250. OP-PM-FP-1001A, Surveillance Test -Monthly Fire Protection System Inspection (Week 1), Revision 25 6.251. OP-PM-FP-1001B, Surveillance Test -Monthly Fire Protection System Inspection (Week 2), Revision 33 6.252. OP-PM-FP-1001C, Monthly Fire Protection System Inspection (Week 3), Revision 13 Page 93 Omaha Public Power District FCS NFPA 805 Transition Report 6.253. OP-PM-FP-1001D, Monthly Fire Protection System Inspection (Week 4), Revision 11 6.254. OP-ST-FP-0001A, Fire Protection System Inspection and Test, Revision 16 6.255. PED-CSS-3, Procuring, Applying and Inspecting Protective Coatings Inside Reactor Containment Building, Revision 7 6.256. PED-ESS-10, Cable Installation, Revision 4 6.257. PED-ESS-17, Conduit Systems, Revision 3 6.258. PED-GEI-4, Fire Protection System Interaction, Revision 6 6.259. PED-QP-1 1, Independent Design Verification (IDV) and Independent Review of Configuration Changes, Revision 10 6.260. PED-QP-3, Calculation Preparation, Review and Approval, Revision 17 6.261. PED-QP-5, Engineering Analysis Preparation, Review and Approval, Revision 28 6.262. PED-SEI-19, System Health Report Preparation, Revision 19 6.263. PED-SEI-28, Program Instructions, Revision 11 6.264. PED-SEI-37, Probabilistic Risk Assessment Configuration Control, Revision 7 6.265. QAP-3.4, Records Management, Revision 10 6.266. QAP-6.7, Fire Protection, Revision 10 6.267. QAP-Appendix F, Fire Protection Equipment in the QA Plan, Revision 4 6.268. SO-C-2, Standing Order -FCS Quality Assurance Records, Revision 92 6.269. SO-G-7, Standing Order -Operating Manual, Revision 68 6.270. SO-G-21, Standing Order -Modification Control, Revision 85 6.271. SO-G-28, Standing Order -Station Fire Plan, Revision 76 6.272. SO-G-30, Standing Order -Procedure Changes and Generation, Revision 121 6.273. SO-G-58, Standing Order -Control Of Fire Protection System Impairments, Revision 36 6.274. SO-G-64, Standing Order -Medical Examination Program For Worker Qualification, Revision 33 6.275. SO-G-74, Standing Order -Fort Calhoun Station EOP/AOP Generation Program, Revision 14 6.276. SO-G-91, Standing Order -Control and Transportation of Combustible Materials, Revision 26 6.277. SO-G-1 02, Standing Order -Fire Protection Program Plan, Revision 8 6.278. SO-G-1 03, Standing Order -Fire Protection Operability and Surveillance Requirements, Revision 25 Page 94 Omaha Public Power District FCS NFPA 805 Transition Report 6.279. SO-G-108, Pollution Prevention and Storm Water Management Plan, Revision 9 6.280. SO-M-9, Standing Order -Hot Work Operations, Revision 28 6.281. SO-O-21, Standing Order -Shutdown Operations Protection Plan, Revision 37 6.282. SO-O-41, Control of Operator Aids and Emergency Equipment, Revision 122 6.283. SO-0-44, Standing Order -Administrative Controls for the Locking of Components, Revision 106 Page 95 Omaha Public Power District FCS NFPA 805 Transition Report Omaha Public Power District FCS NFPA 805 Transition Report Attachments Page 96 Omaha Public Power District FCS NFPA 805 Transition Report B. NEI 04-02.Table B-2 -Nuclear Safety Capability Assessment

-Methodology Review 137 Pages Attached Page B-I Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed.

The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included.

Availability and reliability of equipment selected shall be evaluated.

NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance 3.0 Deterministic Methodology This section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire safe shutdown analysis to address regulatory requirements.

The plant-specific analysis approved by NRC is reflected in the plant's licensing basis.The methodology described in this section is also an acceptable method of performing a post-fire safe shutdown analysis.

This methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships.

This guidance document, however, does not require the use of a computer database oriented approach.The requirements of Appendix R Sections Ilil.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.ADRlicablitv Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036, Section 1.0: Fort Calhoun Station uses a deterministic methodology to assess conformance with the requirements of NFPA 805, Section 4.2, Nuclear Safety. The methodology has been reviewed in detail against NEI 00-01 guidance in the subsequent sections.Comments Reference Document FCS Engineering Analysis EA10-036 Doc, Details Section 1.0 Page B-2 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1 Safe Shutdown Systems and Path Development NEI 00-01 Section 3.0 Guidance This section discusses the identification of systems available and necessary to perform the required safe shutdown functions.

It also provides information on the process for combining these systems into safe shutdown paths. Appendix R Section III.G.1.a requires that the capability to achieve and maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section III.G.l.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.

The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and components remains free of fire damage for a single fire in any single plant fire area. This goal is accomplished by determining those functions important to achieve and maintain hot shutdown.

Safe shutdown systems are selected so that the capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following:

-Reactivity Control-Pressure Control Systems-Inventory Control Systems-Decay Heat Removal Systems-Process Monitoring

-Support Systems* Electrical systems* Cooling systems These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment.

If these functions are preserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated circuits with the potential for spurious equipment operation and/or loss of power source, and the common enclosure failures.

Spurious operations/actuations can affect the accomplishment of the post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious operations of concern are the following:

-A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability

-A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment.

Common power source and common enclosure concerns could also affect these and must be addressed.

Page B-3 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Comments Applicable Alignment Statement Not Required Alignment Basis Generic paragraph.

Detailed alignment discussed in subsequent reference paragraphs.

Comments Reference Document Doc. Details Page B-4 Omaha Public Power District Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection FCS NFPA 805 Transition Report NEI 00-01 Ref 3.1.1 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance The following criteria and assumptions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.APPlicabilitY Comments Applicable Alignment Statement Not Required Alignment Basis Generic paragraph.

Detailed alignment discussed in subsequent reference paragraphs.

Comments Reference Document Page B-5 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3A1A1.1 Criteria/Assumptions A pplicabili Not Applicable NEI 00-01 Section 3.0 Guidance[BWR] GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths For The BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section III.G.1 of 10 CFR 50, Appendix R. Any of the shutdown paths (methods) described in this report are considered to be acceptable methods for achieving redundant safe shutdown.Comments FCS is a PWR; BWR guidance not applicable Alignment Statement Alignment Basis Comments Reference Document Doc, Details Page B-6 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.1.2 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance[BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group (BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown.

The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of 10 CFR 50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000.ApplicabilitY Not Applicable Comments FCS is a PWR; BWR guidance not applicable Alionment Statement Alignment Basis Reference Document Doc, Details Page B-7 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re 3.1.1.3 Criteria/Assumptions ADolicabilitv Applicable Alignment Statement NEI 00-01 Section 3.0 Guidance[PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators.

The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.

Comments Aligns with Intent Alianment Basis Engineering analysis EA10-036, Section 4.4, Item #5: "FCS has design features and procedures to ensure that an adequate source of heat input is maintained for RCS pressure control in sustained Mode 3 (Hot Shutdown Condition)(i.e., a minimum of 150kW of pressurizer heater input to maintain the RCS sub-cooled) utilizing available combinations of the backup pressurizer heaters (banks -1 and -4 are 225kW each, banks -2 and -3 are 150kW each). The backup pressurizer heaters are capable of being energized from emergency diesel generator power. (Note that where the requirement of 150kW of pressurizer heater input capability cannot be met, the loss of backup pressurizer capability is identified as a Variance From the Deterministic Requirements of NFPA 805, Section 4.2.3 [i.e., VFDR].)Comments Reference Document FCS Engineering Analysis EA10-036 Doc. Details Section 4.4, Item #5 Page B-8 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1.1.4 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance The classification of shutdown capability as alternative shutdown is made independent of the selection of systems used for shutdown.Alternative shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path.Compliance to the separation requirements of Sections III.G.1 and II.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.

Aicability Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036, Definition 4.3.1: "Alternate Shutdown Capability

/ Dedicated Shutdown Capability

-NFPA 805 does not provide any definitions of, or make any distinctions for "Alternate Shutdown" or "Dedicated Shutdown" capability.

At FCS, Fire Areas 41 (Cable Spreading Room) and 42 (Main Control Room) are designated as the 10 CFR 50 Appendix R Alternate Shutdown Fire Areas. These two fire areas will transition to NFPA 805 as "Performance Based" due to there being a number of credited Recovery Actions for each area, which are the operator manual actions credited to achieve and maintain "safe and stable" plant operation that are not being performed at a Primary Control Station. The Primary Control Station(s) for FCS are Auxiliary Feedwater Panel A1-179, Alternate Shutdown Al-185, and Electrical Panel AI-212. These three panels are located adjacent to each other in the Upper Electrical Penetration Room (Fire Area 34B-1). The credited Recovery Actions for Fire Areas 41 and 42 are considered by the licensee to have prior NRC approval, consistent with the approved Fire Protection Program per the Fort Calhoun Operating License No. DPR-40, as of Amendment No. 255. The credited Recovery Actions for Fire Areas 41 and 42 have been identified as Variances from Deterministic Requirements (i.e., Variances from the deterministic requirements of NFPA 805 Section 4.2.3) for the purposes of assessing the risk of Recovery Actions.The safe shutdown capabilities provided for the Cable Spreading Room and Main Control Room deterministic fire events include the following:

Decay Heat Removal Capability:

-Align turbine-driven AFW pump FW-1 0 to provide feedwater supply from the EFWST to both Steam Generators through the AFW header, from the Primary Control Station (Al-179) following MCR evacuation and with manual actions to align the flowpath.Process Monitoring Capability:

-Pressurizer Level Channel B (LT-101Y)

@ Al-185-Pressurizer Wide Range Pressure Channel C (PT-1 15) @ Al-1 79-Neutron Flux Wide Range Channel D (NE-004) @ AI-212 Page B-9 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection-RCS Loop 1 Hot Leg Temperature Channel A (TE-l111H)

@ Al-185-RCS Loop 1B Cold Leg Temperature Channel A (TE-1 11 C) @ Al-185-RCS Loop 2 Hot Leg Temperature Channel B (TE-121H)

@ A1-185-RCS Loop 2A Cold Leg Temperature Channel B (TE-121C)

@ A1-185-Steam Generator RC-2A Wide Range Level Channel D (D/LT-91 1) @ Al-1 79-Steam Generator RC-2A Wide Range Pressure Channel D (D/PT-913)

@ A1-179-Steam Generator RC-2B Wide Range Level Channel D (D/LT-912)

@ Al-1 79-Steam Generator RC-2B Wide Range Pressure Channel D (D/PT-914)

@ A1-179-Raw Water Strainer AC-12A Local Pressure Indicator (PI-2805A-1/2) local @ Intake Structure RCS Inventory Control:-Provide borated water from the SIRWT using CVCS pump CH-1B (pump controlled from AI-185, or locally operated at breaker) or CH-1C (pump locally operated at breaker) as a source of RCS inventory makeup. Aligning the RCS makeup path requires local manual action. RCS integrity can be established from the Primary Control Station (Al-185)following MCR evacuation, and with manual actions.RCS Pressure Control:-Initially maintain RCS pressure by operating pressurizer heater RC4-4 locally at the breaker. RCS integrity for inventory control establishes a pressure boundary.Reactivity Control:-Trip the reactor from the MCR prior to evacuation.

Although not required for Hot Shutdown, additional shutdown margin is provided by the injection of borated water from the SIRWT to the RCS using CVCS pump CH-1 B (pump controlled from Al-1 85, or locally operated at breaker) or CH-1 C (pump locally operated at breaker).Vital Auxiliaries:

-Raw Water pump AC-10B or AC-IOD is operated locally at the breaker to support EFWST refill. Manual actions are required to align the Raw Water flowpath.

MCR HVAC is not credited due to MCR evacuation at the onset of the fire.-Vital 4kV bus 1A4 can be aligned with manual actions, powered from onsite (EDG) power." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Definition 4.3.1 Page B- 10 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1.1.5 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown.

Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress.

The units are assumed to be operating at full power under normal conditions and normal lineups.ADDlicabili Comments Applicable Alignment Statement Aligns with Intent Alignment Basis Engineering analysis EA10-036: Assumption 4.2.2: "Plant equipment is assumed to be in its normal expected position or condition at the onset of the fire (with the plant at power operation).

In cases where the status of equipment is indeterminate or could change as a result of expected plant conditions, worst-case initial conditions are assumed for the purpose of cable selection." Assumption 4.2.3: "At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown.

Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Assumptions 4.2.2 and 4.2.3 Page B-1I Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.1.6 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients need be considered in conjunction with the fire.ADRlicabilit Applicable Alianment Statement Aligns Alignment Basis Engineering analysis EA10-036, Assumption 4.2.4: "No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire-induced transients need be considered in conjunction with the fire." Comments Reference Document FCS Engineering Analysis EA10-036 Doc. Details Assumption 4.2.4 Page B-12 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1.1.7 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.ADlicabilit Applicable Alignment Statement Aligns Algnment ai Engineering analysis EA10-036, Assumption 4.2.5: "For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available." The NFPA 805 safe shutdown model includes plant equipment and cables necessary to assess the availability of 161 kV Offsite power for 4kV Switchgear 1A3 and 1A4.Comments Reference Document FCS Engineering Analysis EA10-036 Doc, Details Assumption

4.2.5 Attachment

1 Page B-13 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Rf NEI 00-01 Section 3.0 Guidance 3.1.1.8 Criteria/Assumptions Post-fire safe shutdown systems and components are not required to be safety-related.

Aoplicabilit Comments Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036, Assumption 4.2.6: "The development of the updated NFPA 805 safe shutdown model did not impose any restrictions to only include safety related equipment, or to not include any non-safety related equipment." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Assumption 4.2.6 Page B-14 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance 3.1.1.9 Criteria/Assumptions The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip.

Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis.

At least one train can be repaired or made operable within 72 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.Aoliplcabilit Comments Applicable Alignment Statement Aligns with Intent Alignment Basis Engineering analysis EA10-036: Assumption 4.2.7: "Appendix R post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip.

Under the NFPA 805 Performance-Based Standard, the nuclear safety goal is to provide reasonable assurance that a fire will not prevent the plant from maintaining the fuel in a safe and stable condition.

Consequently, the analysis can demonstrate that the plant can be maintained in hot shutdown until such time that plant systems can be recovered for transition to cold shutdown.

Fire-induced impacts that provide no adverse consequences to achieving and maintaining hot shutdown need not be included in the post-fire safe shutdown analysis.

At least one train can be repaired or made operable using onsite capability to achieve cold shutdown." Section 4.4: "The Nuclear Safety Goal of NFPA 805 is to provide reasonable assurance that a fire during any plant operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a "safe and stable" condition (Ref. Sub-Section 1.3.1 of NFPA 805).Per NFPA 805 the definition of "safe and stable" is (Ref. NFPA 805, definition 1.6.56): Safe and Stable Conditions.

For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff < 0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby (FCS clarification:

the NFPA 805 term "hot standby" for FCS corresponds to Mode 3 [Hot Shutdown Condition], per the FCS Technical Specifications) for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff < 0.99 and fuel coolant temperature below boiling (FCS clarification:

the NFPA 805 term "below boiling" for FCS corresponds to "sub-cooled, not boiling at pressure").(Ref. FCS Technical Specifications, Definitions, as of Amendment No. 265, and EOP-20, "Functional Recovery Procedure")

The NFPA 805 safe shutdown model for FCS has been developed to ensure that the plant can achieve and maintain the fuel in a "safe and stable" condition assuming that a fire event occurs during FCS Mode 1 (Power Operation Condition), Mode 2 (Hot Standby Condition), or Mode 3 (Hot Shutdown Condition).

Page B-15 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection The objective of the deterministic NFPA 805 safe shutdown analysis is, for an all consuming fire occurring in any one plant fire area, to demonstrate that FCS can achieve and maintain Mode 3 (Hot Shutdown Condition), with the minimum plant operating shift staff, for a coping time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.The 24-hour coping time has been selected by FCS based on the design capacity for the backup nitrogen supply that is relied upon to maintain positive remote control over the turbine driven auxiliary feedwater pump, and based on the ability of the FCS Emergency Response Organization to respond to the event, with adequate time allowed for the ERO personnel to muster, assess the extent of fire damage, and assist the plant operating staff with implementation of the required actions to sustain Mode 3 (Hot Shutdown Condition), beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or to assess the extent of fire damage, and assist the plant operating staff with implementation of cold shutdown actions and/or cold shutdown repairs for the plant to transition to, and enter, Mode 4 (Cold Shutdown Condition)." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Assumption

4.2.7 Section

4.4 Page B- 16 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NE00-01 Ref 3.1.1.10 Criteria/Assumptions Aolablial Applicable Alignment Statement Aligns NEI 00-01 Section 3.0 Guidance Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option.Comments Alignment Basis Engineering analysis EA10-036, Assumption 4.2.8: "Manual initiation from the main control room or primary control stations of systems required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option, if the additional cables and equipment are also included in the analysis.

Spurious actuation of automatic systems (Safety Injection, RCS Pressure, etc.) due to fire damage to the input sensors, or due to the likelihood of plant transients that are caused by spurious operation of fire damaged equipment, however, should be evaluated." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Assumption 4.2.8 Page B-17 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re 3.1.1.11 Criteria/Assumptions NEI 00-01 Section 3.0 Guidance Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and maintain safe shutdown for each affected unit must be demonstrated.

Applicability Not Applicable Comments FCS is a single-unit plant; multi-unit guidance not applicable Alignment Statement Alignment Basis Comments Reference Document Doc. Details Page B-18 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.2 Shutdown Functions NEI 00-01 Section 3.0 Guidance The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown.

For additional information on BWR system selection, refer to GE Report GENE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths for the BWR." Applicability Applicable Alignment Statement Comments Not Required Alignment Basis Generic paragraph.

Detailed alignment discussed in subsequent reference paragraphs.

Comments Reference Document Page B-19 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.2.1 Reactivity Control NEI 00-01 Section 3.0 Guidance[BWR] Control Rod Drive System The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability.

Manual scram/reactor trip is credited.

The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.ADolicabilitv Not Applicable Comments FCS is a PWR; BWR guidance not applicable Alignment Statement Alionment Basis Comments Reference Document Page B-20 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 RQf NEI 00-01 Section 3.0 Guidance 3.1.2.1 Reactivity Control [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.

Aoolicability Comments Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036: Section 4.4, Item #3: "The FCS reactor core design ensures that Keff is maintained

<0.99 while the plant is in sustained Mode 3 (Hot Shutdown Condition).

Consequently, maintaining the "safe and stable" plant condition for NFPA 805 will not require boration of the RCS. Gravity insertion of the control rods into the reactor core will ensure reactivity control is achieved and maintained for Mode 3 (Hot Shutdown Condition)." Section 4.4, Item #4: "Inventory makeup to the RCS may only be required to account for expected RCS leakage and minimal RCS shrinkage.

FCS has design features and procedures to ensure that an adequate source of borated inventory is provided for RCS inventory control in sustained Mode 3 (Hot Shutdown Condition) (i.e., RCS inventory makeup from the SIRWT and/or the BASTs to maintain the RCS sub-cooled) utilizing the CVCS or the HPSI system." Attachment 1, Performance Goal "Reactivity Control": "Reactivity control is maintained by controlling both negative and positive reactivity addition to the Reactor Coolant System (RCS). Negative reactivity is ensured through short term Reactor trip and long term boration of the RCS, although long term boration is not required to ensure reactivity control for Hot Shutdown.

Control over positive reactivity is ensured by isolating potential boron dilution flowpaths, and by preventing uncontrolled RCS temperature decrease." Attachment 1, Performance Goal "Reactivity Control": "Although not required for Hot Shutdown, additional shutdown margin (negative reactivity) for Cold Shutdown is assured through the use of borated water for RCS makeup. The Chemical and Volume Control System (CVCS) and High Pressure Safety Injection System (HPSI) are modeled under Performance Goal RCS-INV (see below [reference within engineering analysis EA10-036])

for inventory control purposes, and are credited to provide only borated sources of water for RCS inventory control from the Boric Acid Storage Tanks (BASTs), or the Safety Injection Refueling Water Storage Tank (SIRWT)." Page B-21 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Comments Reference Document FCS Engineering Analysis EA10-036 Doc, Details Section 4.4, Items #3 and #4 Attachment 1, Performance Goal "Reactivity Control" Page B-22 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.2.2 Pressure Control Systems NEI 00-01 Section 3.0 Guidance[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.ApolicabilitY Not Applicable FCS is a PWR; BWR guidance not applicable Alignment Statement Aliganment Basis Comments Reference Document Page B-23 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re NEI 00-01 Section 3.0 Guidance 3.1.2.2 Pressure Control Systems [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.

Agalicabii Comments Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036, Attachment 1, Performance Goal "Inventory and Pressure Control": "RCS pressure and inventory is controlled by ensuring that all potential pressure boundary and flow diversion flowpaths from the RCS are isolated, and by providing the Pressurizer Heaters to maintain RCS pressure, and inventory makeup capability from CVCS and/or HPSI Systems to maintain RCS inventory.

Reactor sub-cooling is maintained by establishing a controlled rate of RCS cooldown utilizing natural circulation and rejecting core decay heat to either of the two Steam Generators (see Decay Heat Removal, below [reference within engineering analysis EA10-036]).

RCS inventory loss is controlled by modeling isolation of RCS pressure boundaries:

Pressurizer Power Operated Relief Valves (PORVs), Reactor Coolant Gas Vent System (RCGVS) Valves and RCS Letdown Valves.RCS inventory makeup water is modeled by injecting water from the BASTs or the SIRWT to the RCS via the Main Charging Header. Although not required for Hot Shutdown, this is also providing a means of long term reactivity control, injecting borated water from the SIRWT or BASTs (see Reactivity Control, above [reference within engineering analysis EA10-036]).

Alternate Charging paths are available to provide makeup water by utilizing the HPSI Header. The HPSI Header can be supplied from either the CVCS or the HPSI Pumps.Note that the normal and alternate Charging paths are not credited for once through cooling (e.g. "feed and bleed") in the safe shutdown analysis; decay heat removal is always provided by the Auxiliary Feedwater System (AFW) (see Decay Heat Removal, below [reference within engineering analysis EA10-036]).

Pressure control is accomplished primarily by controlling RCS makeup and RCS cooldown rates. Spurious operations are modeled that could interfere with RCS pressure control (i.e., PORVs opening, RCGVS Valves opening, Pressurizer Main Spray Valves opening, Pressurizer Auxiliary Spray Valves opening, Reactor Coolant Pumps [RCPs]running, non-credited Pressurizer Heaters energized, non-credited Charging Pumps running, etc.). Note that Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that Hot Shutdown can be maintained without the use of Pressurizer Heaters. However, the Pressurizer Heaters are modeled for the Inventory and Pressure Control Performance Goal since they are considered to be an aid to the plant operator for maintaining RCS pressure.Page B-24 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection RCS pressure reduction can be achieved using Pressurizer vent paths, or by utilizing Pressurizer Auxiliary Spray." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Attachment 1, Performance Goal "Inventory and Pressure Control" Page B-25 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref 3.1.2.3 Inventory Control Aoplicabilitv Not Applicable Alianment Statement Alignment Basis Comments Reference Document NEI 00-01 Section 3.0 Guidance[BWR] Systems selected for the inventory control function should be capable of supplying sufficient reactor coolant to achieve and maintain hot shutdown.

Manual initiation of these systems is acceptable.

Automatic initiation functions are not required.Comments FCS is a PWR; BWR guidance not applicable Page B-26 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1.2.3 Inventory Control ADplicabili Applicable Alignment Statement Aligns NEI 00-01 Section 3.0 Guidance[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.Comments Alignment Basis Engineering analysis EA10-036, Attachment 1, Performance Goal "Inventory and Pressure Control": "RCS pressure and inventory is controlled by ensuring that all potential pressure boundary and flow diversion flowpaths from the RCS are isolated, and by providing the Pressurizer Heaters to maintain RCS pressure, and inventory makeup capability from CVCS and/or HPSI Systems to maintain RCS inventory.

Reactor sub-cooling is maintained by establishing a controlled rate of RCS cooldown utilizing natural circulation and rejecting core decay heat to either of the two Steam Generators (see Decay Heat Removal, below [reference within engineering analysis EA10-036]).

RCS inventory loss is controlled by modeling isolation of RCS pressure boundaries:

Pressurizer Power Operated Relief Valves (PORVs), Reactor Coolant Gas Vent System (RCGVS) Valves and RCS Letdown Valves.RCS inventory makeup water is modeled by injecting water from the BASTs or the SIRWT to the RCS via the Main Charging Header. Although not required for Hot Shutdown, this is also providing a means of long term reactivity control, injecting borated water from the SIRWT or BASTs (see Reactivity Control, above [reference within engineering analysis EA1 0-036]).Alternate Charging paths are available to provide makeup water by utilizing the HPSI Header. The HPSI Header can be supplied from either the CVCS or the HPSI Pumps.Note that the normal and alternate Charging paths are not credited for once through cooling (e.g. "feed and bleed") in the safe shutdown analysis; decay heat removal is always provided by the Auxiliary Feedwater System (AFW) (see Decay Heat Removal, below [reference within engineering analysis EA10-036])." Comments Reference Document Doc. Details FCS Engineering Analysis EA10-036 Attachment 1, Performance Goal "Inventory and Pressure Control" Page B-27 Omaha Public Power District Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection FCS NFPA 805 Transition Report NEI00-01 Rf 3.1.2.4 Decay Heat Removal ARlicability Not Applicable Alignment Statement NEI 00-01 Section 3.0 Guidance[BWR] Systems selected for the decay heat removal function(s) should be capable of:-Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure.-Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool).-Removing sufficient decay heat from the reactor to achieve cold shutdown.This does not restrict the use of other systems.Comments FCS is a PWR; BWR guidance not applicable Alianment Basis Comments Reference Document Page B-28 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re NEI 00-01 Section 3.0 Guidance 3.1.2.4 Decay Heat Removal [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.Aolicabiliv Comments Applicable Alignment Statement Aligns with Intent Alianment Basis Engineering analysis EA10-036: Section 4.4, Item #1: "FCS has design features and procedures to ensure that an adequate source of inventory is provided for decay heat removal in sustained Mode 3 conditions (i.e., EFWST re-fill capability from raw water for the motor driven AFW pump and the turbine driven AFW pump [Ref. AOP-30, "Emergency Fill of Emergency Feedwater Storage Tank"], and/or alternate water supply for the diesel driven AFW pump from the main condenser upon depletion of the CST)." Section 4.4, Item #2: "Core decay heat in the Mode 3 (Hot Shutdown Condition) will be rejected to the secondary plant through one or both of the steam generators, and then to atmosphere through the main steam safety relief valves operating as spring relief valves." Attachment 1, Performance Goal "Decay Heat Removal": "Core decay heat removal while in Hot Shutdown is primarily accomplished by dumping steam from at least one Steam Generator, using at least one of three AFW Pumps and the Emergency Feedwater Storage Tank (EFWST) or Condensate Storage Tank (CST) for makeup water. Long term refill of the EFWST is provided by the Raw Water System (RW) in accordance with procedure AOP-30, "Emergency Fill of Emergency Feedwater Storage Tank", while alternate water supply to Engine Driven AFW Pump FW-54 is provided from the Condenser upon depletion of the CST." Page B-29 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Comments Reference Document FCS Engineering Analysis EA1 0-036 Doc. Details Section 4.4, Items #1 and #2 Attachment 1, Performance Goal "Decay Heat Removal" Page B-30 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection 3.1.2.5 Process Monitoring NEI 00-01 Section 3.0 Guidance The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1,Section IX "Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10 CFR 50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function.

This instrumentation is that which monitors the process variables necessary to perform and control the functions specified in Appendix R Section II.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3).

IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (III.G.1 and lIl.G.2).

In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures (including Abnormal Operating Procedures).

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 system 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.ADUlicability Comments Applicable Alianment Statement Aligns with Intent Alianment Basis Engineering analysis EA10-036, Attachment 1, Performance Goal "Process Monitoring": "NEI 00-01 contains guidance in Section 3.1.2.5 for choosing appropriate instrumentation for process monitoring.

In general, instruments required for the safe shutdown model are directly associated with the component, system, or performance goal that the instrument supports: Reactor Coolant Hot and Cold Leg Temperature:

These instruments are modeled in support of Process Monitoring""Pressurizer Pressure and Level: These instruments are modeled in support of Process Monitoring""Neutron Flux Monitoring (source range): These instruments are modeled in support of Reactivity Control" Page B-31 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection"Level Instrumentation for various tanks: Level indication is provided for the EFWST to provide indication to the Main Control Room when used as a source for AFW makeup supply.Level indication has not been included for a number of credited tanks, including the SIRWT and BASTs. Plant calculation FC06718, "Appendix R -Basis for Tank Level Instrumentation

/ Raw Water Pump Component Selection", has been completed to show that adequate shutdown margin and inventory is available to shutdown the plant. The calculation confirms that adequate shutdown inventory will be available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following plant shutdown.""Level indication has not been included for the Condensate Storage Tank (CST), DW-48, the Diesel Generator Fuel Oil Storage Tank, FO-1, and the Auxiliary Boiler Fuel Oil Tank, FO-10 (the fuel oil supply for Diesel Driven AFW Pump FW-54 and an additional source of supply for the Emergency Diesel Generators).

The levels of Tanks FO-1 and FO-10 are controlled by FCS Technical Specification 2.7 to maintain 7 days of fuel oil available on site for continuous operation of one Emergency Diesel Generator.

The minimum fuel oil inventory is adequate for approximately 5 days of continuous operation with one Emergency Diesel Generator running at full load (4,400 gallons per day) and FW-54 running (480 gallons per day). The bases for FCS Technical Specification

2.7 identify

that plant procedures and methods have been established for the transfer of fuel oil from FO-10 to FO-1 as necessary to prevent interruption of Emergency Diesel Generator function with the onsite fuel oil supply, and that 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is a sufficient amount of time for offsite sources of fuel oil to be provided to the site and analyzed prior to FO-1 and/or FO-10 replenishment.

The CST is administratively controlled to maintain tank level at no less than 50%, representing 75,000 gallons of inventory, based on Annunciator Response Procedure ARP-CB-10, 11/All, B6-L window. AFW Pump FW-54 can be provided with an alternate source of makeup water from the Condenser should the CST inventory become depleted.""For the Emergency Diesel Generators, the NFPA 805 safe shutdown model includes automatic fuel makeup capability from the Diesel Generator Fuel Oil Storage Tank, FO-1, inclusive of Auxiliary Fuel Oil Tank(s) level instrumentation and controls.Diesel Driven AFW Pump FW-54 includes automatic fuel oil makeup from FO-10 when level in the diesel (FW-56) day tank reaches the low set point as sensed by LS-2120. A minimum volume of 8,000 gallons of diesel fuel oil is maintained in FO-10 by FCS Technical Specification 2.7. The NFPA 805 safe shutdown model includes the automatic fuel makeup capability for the Fuel Oil Day Tank (Pump FO-37, Level Switch LS-2120, and associated power supplies).

The diesel day tank is sized for approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation (160 gallons of diesel fuel oil) and is provided with low level annunciation in the Main Control Room. The diesel day tank level is administratively monitored based on Annunciator Response Procedure ARP-CB-10,1 1A21, B5-L window.FCS Technical Specification

2.7 ensures

that the Emergency Diesel Generators will have an adequate supply of diesel fuel oil should Offsite power not be available due to fire damage. FCS Technical Specification 2.7 and administrative controls ensure that the FW-54 will have an adequate supply of AFW makeup water and diesel fuel oil should FW-54 be the credited AFW Pump for shutdown.

Furthermore, for tanks that are not provided with level monitoring instrumentation, the safe shutdown analysis models spurious flow diversion paths, as applicable, to identify fire areas where inadvertent tank draindown could potentially occur (due to pumped diversion or by gravity drain).""Steam Generator Level and Pressure:

These instruments are modeled in support of Decay Heat Removal""Diagnostic Instrumentation for safe shutdown systems: Diagnostic instrumentation such as pump discharge pressure, flow and temperature are generally available from local indicators that require no electrical power. Where beneficial to reduce operator burden, instruments that read out in the Main Control Room have been included in the model and logically associated with the component being monitored.

The Raw Water Pumps are supported by Raw Water local Strainer Differential Pressure Instruments to provide indication to operators of strainer performance.

Similarly, the Motor Driven and Turbine Driven AFW Pumps are supported by EFWST Level Indication." Page B-32 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Comments Reference Document FCS Engineering Analysis EA10-036 Doc. Details Attachment 1, Performance Goal "Process Monitoring" Page B-33 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI00-01 Ref NEI 00-01 Section 3.0 Guidance 3.1.2.6.1 Electrical Systems AC Distribution System Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1 E busses either directly from the busses or through step down transformers/

load centers/ distribution panels for 600, 480 or 120 VAC loads.For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sources or the emergency diesel generator depending on which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a loss of offsite power. Refer to Section 3.1.1.7.DC Distribution System Typically, the 125VDC distribution system supplies DC control power to various 125VDC control panels including switchgear breaker controls.

The 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters.

These distribution panels typically supply power for instrumentation necessary to complete the process monitoring functions.

For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the diesel generators to become operational.

Once the diesels are operational, the 125 VDC distribution system can be powered from the diesels through the battery chargers.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).ADDlicabilitv Comments Applicable Alianment Statement Aligns Alignment Basis Engineering analysis EA10-036, Attachment 1, System "Electrical Distribution Systems": "The following is a high level summary of the NFPA 805 safe shutdown model for the Electrical Distribution Systems/Component at FCS:-161kV Switchyard

1. 1251, Breakers 1251-110 and 1251-111, 161/4kV Transformers T1A-3 and T1A-4, Motor Operated Disconnects DS-T1A-3 and DS-T1A-4 Page B-34 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection-4kV Switchgear 1A3 and 1A4 supplied from either 161kV Offsite Power (via Transformers TIA-3 and T1A-4) or the associated Emergency Diesel Generator (DG-1 or DG-2)-480V Switchgear 1 B3A, 1 B3A-4A, 11B3B, 1 B3B-4B, 1 B3C, 1 B3C-4C, 1 B4A, 11B4B, and 1 B4C-480V Motor Control Centers 3A1 /1 B3A-2, 3A2/1 B3A-3, 3A4/1 B3A-6, 3B1/11B3B-2, 3B3/11B3B-6, 3C 1/1 B3C-1, 3C2/1 B3C-2, 4A1 /1 B4A-2, 4A2/1 B4A-3, 4B1/11B4B-2, 4B2/1 B4B-5, 4C1 /1 B4C-2, 4C2/1 B4C-3, 4C4/1 B4C-7, and MCC-4C6-480V AC Transformer Auxiliary Power Panels MPP-69 and MPP-70-125V DC Station Batteries EE-8A and EE-8B-125V DC Battery Chargers EE-8C, EE-8D, EE-8E, and EE-57-125V DC Main Distribution Panels EE-8F and EE-8G-125V DC Buses AI-41A and AI-41B-125V DC Panel DC-PNL-1-120V AC Instrument Bus Inverters EE-8H, EE-8J, EE-8K, EE-8L, EE-8P, EE-8Q, EE-8T, and EE-8U-120V AC Instrument Buses AI-40A, AI-40B, AI-40C, AI-40D, AI-42A, and AI-42B" Per the FCS PRA Summary Notebook, Rev. 12, July 29, 2010, Section 6.3.10, the station batteries can supply plant loads for a minimum of 2.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following loss of AC power or loss of the associated battery chargers, assuming failure of the plant operators to shed non-essential DC loads. The capacity of the station batteries provides sufficient margin for the plant operators to be able to establish a means of battery charging prior to battery depletion.

Comments Reference Document Doc Details FCS Engineering Analysis EA10-036 Attachment 1, System "Electrical Distribution Systems" FCS PRA Summary Notebook Section 6.3.10 Page B-35 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance 3.1.2.6.2 Cooling Systems Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations.

Typical uses include:-RHR/SDC/DH Heat Exchanger cooling water-Safe shutdown pump cooling (seal coolers, oil coolers)-Diesel generator cooling-HVAC system cooling water.HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer's literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents). HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations.

Typical uses include:-Main control room, cable spreading room, relay room-ECCS pump compartments

-Diesel generator rooms-Switchgear rooms Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.

AnR1licabkili Comments Applicable Alignment Statement Aligns Alignment Basis Engineering analysis EA10-036, Attachment 1: "The Raw Water System provides support for Control Room HVAC, SDC Heat Exchangers, the LPSI Pump Coolers, as well as the ability to refill the EFWST for long term decay heat removal.""Main Control Room ventilation is modeled in the safe shutdown analysis to maintain Main Control Room habitability." Page B-36 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Comments Section 3.1.2.6.2 of NEI 00-01 Guidance states that plant-specific evaluations are necessary to determine the HVAC systems essential to support safe shutdown.FC06237 documents the acceptability of limited HVAC requirements (i.e., only MCR HVAC required).

Reference Document Doc, Details FCS Engineering Analysis EA10-036 Attachment 1, Performance Goal "Vital Auxiliaries" FCS Calculation FC06237 Page B-37 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re 3.1.3 Methodology for Shutdown System Selection NEI 00-01 Section 3.0 Guidance Refer to NEI-00-01 Rev 1 Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths.The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis: ADi~licabili Comments Applicable Alignment Statement Not Required Alignment Basis Generic paragraph.

Detailed alignment discussed in subsequent reference paragraphs.

Comments Reference Document Page B-38 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Re NEI 00-01 Section 3.0 Guidance 3.1.3.1 Identify Safe Shutdown Review available documentation to obtain an understanding of the available plant systems and the functions required to achieve and Functions 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" Apakliy Comments Applicable Alignment Statement Aligns Alianment Basis Engineering analysis EA10-036, Attachment 14, Section 5.1.1: "The systems engineer will review the following documentation as applicable to perform this selection:-Operating Procedures-Piping and Instrumentation Diagrams (P&IDs)-Appendix R Logic diagrams-Electrical distribution 1-line Diagrams-System Descriptions-System Design Basis Documents-Licensing Documents-Calculations" Engineering analysis EA10-036, Attachment 14 (the procedure for post fire safe shutdown/Fire PRA component identification), specifies the documents that shall be reviewed to determine the required plant systems for safe shutdown.

In addition, Section 5.0 of engineering analysis EA10-036 contains a list of specific references that were reviewed, and Attachment 1 of EA1 0-036 identifies references used to review the NFPA 805 safe shutdown model. It is implicit that the content of these references was reviewed as applicable to the identification of safe shutdown functions.

Page B-39 Omaha Public Power District FCS NFPA 805 Transition Report Table B-2 Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.1 Nuclear Safety Capability System and Equipment Selection Reference Document FCS Engineering Analysis EA10-036 Doc. Details Section 5.0 Attachment 1 Attachment 14, Section 5.1.1 Page B-40